Study of plastic solidification process on solid radioactive waste treatment

International Nuclear Information System (INIS)

Jing Weiguan; Zhang Yinsheng; Qian Wenju

1994-01-01

Comparisons between the plastic solidification conditions of incinerated ash and waste cation resin by using thermosetting plastic polyvinyl chloride (PVC), polystyrene (PS) and polyethylene (PE), and identified physico-chemical properties and irradiation resistance of solidified products were presented. These solidified products have passed through different tests as compression strength, leachability, durability, stability, permeability and irradiation resistance (10 6 Gy) etc. The result showed that the solidified products possessed stable properties and met the storage requirement. The waste tube of radioimmunoassay, being used as solidification medium to contain incinerated ash, had good mechanical properties and satisfactory volume reduction. This process may develop a new way for disposal solid radioactive waste by means of re-using waste

Improved cement solidification of low and intermediate level radioactive wastes

International Nuclear Information System (INIS)

1993-01-01

Cementation was the first and is still the most widely applied technique for the conditioning of low and intermediate level radioactive wastes. Compared with other solidification techniques, cementation is relatively simple and inexpensive. However, the quality of the final cemented waste forms depends very much on the composition of the waste and the type of cement used. Different kinds of cement are used for different kinds of waste and the compatibility of a specific waste with a specific cement type should always be carefully evaluated. Cementation technology is continuously being developed in order to improve the characteristics of cemented waste in accordance with the increasing requirements for quality of the final solidified waste. Various kinds of additives and chemicals are used to improve the cemented waste forms in order to meet all safety requirements. This report is meant mainly for engineers and designers, to provide an explanation of the chemistry of cementation systems and to facilitate the choice of solidification agents and processing equipment. It reviews recent developments in cementation technology for improving the quality of cemented waste forms and provides a brief description of the various cement solidification processes in use. Refs, figs and tabs

Solidification of radioactive waste resins using cement mixed with organic material

Energy Technology Data Exchange (ETDEWEB)

Laili, Zalina, E-mail: liena@nm.gov.my [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Wahab, Mohd Abdul [Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia)

2015-04-29

Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins.

Solidification of radioactive waste resins using cement mixed with organic material

International Nuclear Information System (INIS)

Laili, Zalina; Yasir, Muhamad Samudi; Wahab, Mohd Abdul

2015-01-01

Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins

Application of sulfur concrete for solidification of radioactive wastes and building of repositories

International Nuclear Information System (INIS)

Cholerzynski, A.; Tomczak, W.; Switalski, J.

2000-01-01

The application of sulfur concrete as solidification material for radioactive wastes and as building material used in repositories have been presented. Their high shear strength, low level of leaching, and high radiation resistance decide of positive recommendation of such material for wide use in radioactive waste treatment processes and repositories building

Alternative method of solidification for low-level class a radioactive waste

International Nuclear Information System (INIS)

Mayo, K.S.

1988-01-01

New solidification media have been developed that exhibit excellent spatial efficiency over the entire range of virtually all Class A liquid wastes. These new media are being used to incorporate from 41 to 48 gallons of liquid radioactive waste in a 55-gallon drum. To date, wastes processing at nuclear power plants and facilities include oils, evaporator bottoms, sludges, and ion-exchanges resins as well as combinations of these waste streams. This paper comparatively discusses the performance of solidification agents known as AQUASET TM and PETROSET TM with other currently available agents. It presents key advantages of using the AQUASET and PETROSET media over other media. These advantages include improvements in packaging efficiency, leachability, and repeatability

Solidification Technologies for Radioactive and Chemical Liquid Waste Treatment - Final CRADA Report

International Nuclear Information System (INIS)

Castiglioni, Andrew J.; Gelis, Artem V.

2016-01-01

This project, organized under DOE/NNSA's Global Initiatives for Proliferation Prevention program, joined Russian and DOE scientists in developing more effective solidification and storage technologies for liquid radioactive waste. Several patent applications were filed by the Russian scientists (Russia only) and in 2012, the technology developed was approved by Russia's Federal State Unitary Enterprise RADON for application throughout Russia in cleaning up and disposing of radioactive waste.

Characteristics of Cement Solidification of Metal Hydroxide Waste

Directory of Open Access Journals (Sweden)

Dae-Seo Koo

2017-02-01

Full Text Available To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

Characteristics of cement solidification of metal hydroxide waste

Energy Technology Data Exchange (ETDEWEB)

Koo, Dae Seo; Sung, Hyun Hee; Kim, Seung Soo; Kim, Gye Nam; Choi, Jong Won [Dept. of Decontemination Decommission Technology Development, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2017-02-15

To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

Method of cement-solidification of radioactive liquid wastes containing surfactant

Energy Technology Data Exchange (ETDEWEB)

Sugimoto, Y; Yusa, H

1979-04-10

Purpose: To provide the subject method comprising the steps of adjusting the concentration of the surfactant to a value less than the predetermined value even when the concentration of the surfactant is high, and rendering the uniaxial compression strength of the cement-solidification body into more than the defined fabrication reference value. Method: To radioactive liquid wastes there are applied means for boiling and heating liquid wastes by addition of sulfuric acid, means for cracking surfactants by the addition of oxidants and means for precipitating and arresting surfactants. After suppressing the hindrance of the cement hydration reaction by surfactants, the radioactive liquid wastes are cement-solidified. (Nakamura, S.).

Solidification of radioactive wastes with inorganic binders (literature survey)

International Nuclear Information System (INIS)

Rudolph, G.; Koester, R.

A survey is provided on solidification of radioactive waste solutions, sludges and tritium waste water through cement and other inorganic binders. A general survey of the possibilities described in the literature is followed by a somewhat more detailed description of the work carried on at four research establishments in the United States, Oak Ridge National Laboratory, Savannah River Laboratory, Brookhaven National Laboratory, and Atlantic Richfield Hanford Company, supplemented by personal information. Additional sections describe the experiences with various types of cement and the possibilities for improvement of solidification products through preliminary fixation of the toxic nuclides (transformation into insoluble products or absorption); there is a further possibility of post-treatment through polymer impregnation. Finally, definition and determination of leachability are provided and some results compiled. 74 references, 7 figures, 5 tables

Small-scale demonstration of high-level radioactive waste processing and solidification using actual SRP waste

International Nuclear Information System (INIS)

Okeson, J.K.; Galloway, R.M.; Wilhite, E.L.; Woolsey, G.B.; Ferguson, R.B.

1980-01-01

A small-scale demonstration of the high-level radioactive waste solidification process by vitrification in borosilicate glass is being conducted using 5-6 liter batches of actual waste. Equipment performance and processing characteristics of the various unit operations in the process are reported and, where appropriate, are compared to large-scale results obtained with synthetic waste

Recent advances in cement solidification of radioactive wastes

International Nuclear Information System (INIS)

Vigreux, B.; Jaouen, C.

1987-01-01

Advanced cement solidification processes and systems have been developed by SGN to meet changing requirements in radioactive waste processing and packaging and to avoid the difficulties often encountered in waste concreting on an industrial scale. SGN applies a strict development methodology to ensure integration of the most recent information on chemical behavior of solidified wastes plus compliance with the precise needs of waste producers and evolving regulatory requirements concerning waste package storage and disposal. Based on a hierarchical definition of objectives, this methodology was implemented following an overall study on radwaste concreting performed in 1983 and 1984 for Electricite de France (EdF), France's national electric power utility. It ensures that industrial and regulatory factors are fully considered from the start of development work. It also constrains development in the direction of true process optimization and guarantees compliance with defined objectives. The methodology has helped SGN develop concreting processes adapted to various types of radioactive waste. The most widely employed processes are first briefly described in this paper. It then presents continuous and batch systems using these processes, focusing on technological features chosen at a very early stage in development

Solidification of low-level radioactive wastes in masonry cement

International Nuclear Information System (INIS)

Zhou, H.; Colombo, P.

1987-03-01

Portland cements are widely used as solidification agents for low-level radioactive wastes. However, it is known that boric acid wastes, as generated at pressurized water reactors (PWR's) are difficult to solidify using ordinary portland cements. Waste containing as little as 5 wt % boric acid inhibits the curing of the cement. For this purpose, the suitability of masonry cement was investigated. Masonry cement, in the US consists of 50 wt % slaked lime (CaOH 2 ) and 50 wt % of portland type I cement. Addition of boric acid in molar concentrations equal to or less than the molar concentration of the alkali in the cement eliminates any inhibiting effects. Accordingly, 15 wt % boric acid can be satisfactorily incorporated into masonry cement. The suitability of masonry cement for the solidification of sodium sulfate wastes produced at boiling water reactors (BWR's) was also investigated. It was observed that although sodium sulfate - masonry cement waste forms containing as much as 40 wt % Na 2 SO 4 can be prepared, waste forms with more than 7 wt % sodium sulfate undergo catastrophic failure when exposed to an aqueous environment. It was determined by x-ray diffraction that in the presence of water, the sulfate reacts with hydrated calcium aluminate to form calcium aluminum sulfate hydrate (ettringite). This reaction involves a volume increase resulting in failure of the waste form. Formulation data were identified to maximize volumetric efficiency for the solidification of boric acid and sodium sulfate wastes. Measurement of some of the waste form properties relevant to evaluating the potential for the release of radionuclides to the environment included leachability, compression strengths and chemical interactions between the waste components and masonry cement. 15 refs., 19 figs., 9 tabs

Survey of agents and techniques applicable to the solidification of low-level radioactive wastes

International Nuclear Information System (INIS)

Fuhrmann, M.; Neilson, R.M. Jr.; Colombo, P.

1981-12-01

A review of the various solidification agents and techniques that are currently available or potentially applicable for the solidification of low-level radioactive wastes is presented. An overview of the types and quantities of low-level wastes produced is presented. Descriptions of waste form matrix materials, the wastes types for which they have been or may be applied and available information concerning relevant waste form properties and characteristics follow. Also included are descriptions of the processing techniques themselves with an emphasis on those operating parameters which impact upon waste form properties. The solidification agents considered in this survey include: hydraulic cements, thermoplastic materials, thermosetting polymers, glasses, synthetic minerals and composite materials. This survey is part of a program supported by the United States Department of Energy's Low-Level Waste Management Program (LLWMP). This work provides input into LLWMP efforts to develop and compile information relevant to the treatment and processing of low-level wastes and their disposal by shallow land burial

Survey of agents and techniques applicable to the solidification of low-level radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Fuhrmann, M.; Neilson, R.M. Jr.; Colombo, P.

1981-12-01

A review of the various solidification agents and techniques that are currently available or potentially applicable for the solidification of low-level radioactive wastes is presented. An overview of the types and quantities of low-level wastes produced is presented. Descriptions of waste form matrix materials, the wastes types for which they have been or may be applied and available information concerning relevant waste form properties and characteristics follow. Also included are descriptions of the processing techniques themselves with an emphasis on those operating parameters which impact upon waste form properties. The solidification agents considered in this survey include: hydraulic cements, thermoplastic materials, thermosetting polymers, glasses, synthetic minerals and composite materials. This survey is part of a program supported by the United States Department of Energy's Low-Level Waste Management Program (LLWMP). This work provides input into LLWMP efforts to develop and compile information relevant to the treatment and processing of low-level wastes and their disposal by shallow land burial.

Toxic and hazardous waste disposal. Volume 1. Processes for stabilization/solidification

International Nuclear Information System (INIS)

Pojasek, R.B.

1979-01-01

Processes for the stabilization and/or solidification of toxic, hazardous, and radioactive wastes are reviewed. The types of wastes classified as hazardous are defined. The following processes for the solidification of hazardous wastes are described: lime-based techniques; thermoplastic techniques; organic polymer techniques; and encapsulation. The following processes for the solidification of high-level radioactive wastes are described: calcination; glassification; and ceramics. The solidification of low-level radioactive wastes with asphalt, cement, and polymeric materials is also discussed. Other topics covered include: the use of an extruder/evaporator to stabilize and solidify hazardous wastes; effect disposal of fine coal refuse and flue gas desulfurization slurries using Calcilox additive stabilization; the Terra-Tite Process; the Petrifix Process; the SFT Terra-Crete Process; Sealosafe Process; Chemfix Process; and options for disposal of sulfur oxide wastes

Disposal of radioactive wastes

International Nuclear Information System (INIS)

Dlouhy, Z.

1982-01-01

This book provides information on the origin, characteristics and methods of processing of radioactive wastes, as well as the philosophy and practice of their storage and disposal. Chapters are devoted to the following topics: radioactive wastes, characteristics of radioactive wastes, processing liquid and solid radioactive wastes, processing wastes from spent fuel reprocessing, processing gaseous radioactive wastes, fixation of radioactive concentrates, solidification of high-level radioactive wastes, use of radioactive wastes as raw material, radioactive waste disposal, transport of radioactive wastes and economic problems of radioactive wastes disposal. (C.F.)

UJV line for research into radioactive wastes solidification

International Nuclear Information System (INIS)

Neumann, L.; Feist, I.; Kepak, F.; Nachmilner, L.; Napravnik, J.; Novak, M.; Pecak, V.; Vojtech, O.

1985-01-01

An experimental line with a capacity of 0.01 m 3 /h was developed and built for research of the solidification of liquid radioactive wastes at the Nuclear Research Institute. The line allows the research and pilot plant testing of processes based on vitrification but also on other procedures including calcination. It consists of a horizontal calciner, a resistance melting unit, a homogenization device for research into cementation of the calcinate, and equipment for the disposal of gaseous emissions. The facility is provided with a control console which allows remote control and the control of all basic operating parameters. The design of the line allows its eventual completion with other equipment. (Z.M.)

Solidification of radioactive waste solutions by pelletization technique

International Nuclear Information System (INIS)

Akbar, A.H.; Koester, R.; Rudolph, G.

1980-04-01

A possible way of performing the cement fixation of radioactive wastes is the incorporation into cement pellets on a pan pelletizer, followed by embedding the pellets into an inactive cement matrix. This procedure is suitable for various types of waste, particularly for medium level liquid wastes, and can be used both at drum disposal and at in-situ solidification. This report describes some initial studies on the pelletization technique using a laboratory pelletizer. Formation and size of the pellets have been found to be determined by speed, angle, and load of the pan, ratio and mode of addition of the liquid and solid components, ect. Pellets in various compositions have been produced from cement and water or simulated waste solution, in some cases with the addition of bentonite for improving cesium retention. Some mechanical properties of the pellets such as fall height of fresh pellets, development of hardness (crush test), impact and abrasion resistance, have been determined. Some preliminary experiments were done on backfilling the void space between the pellets - about 40 per cent of the bulk volume - with cement grouts of appropriate compositions. (orig.) [de

Volume reduction and solidification of liquid and solid low-level radioactive waste

International Nuclear Information System (INIS)

May, J.R.

1979-01-01

This paper presents a brief background of the development of a method of radioactive waste volume reduction using a unique fluidized bed calciner/incinerator. The volume reduction system is capable of processing a variety of liquid chemical wastes, spent ion exchange resin beads, filter treatment sludges, contaminated lubricating oils, and miscellaneous combustible solids such as paper, rags, protective clothing, wood, etc. All of these wastes are processed in one chemical reaction vessel. Detailed process data is presented that shows the system is capable of reducing the total volume of disposable radioactive waste generated by light water reactors by a factor of 10. Equally important to reducing the volume of power reactor radwaste is the final form of the stored or disposable radwaste. This paper also presents process data related to a new radwaste solidification system, presently being developed, that is particularly suited for immobilizing the granular solids and ashes resulting from volume reduction by calcination and/or incineration

Economic analysis of a volume reduction/polyethylene solidification system for low-level radioactive wastes

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1985-01-01

A study was conducted at Brookhaven National Laboratory to determine the economic feasibility of a fluidized bed volume reduction/polyethylene solidification system for low-level radioactive wastes. These results are compared with the ''null'' alternative of no volume reduction and solidification of aqueous waste streams in hydraulic cement. The economic analysis employed a levelized revenue requirement (LRR) technique conducted over a ten year period. An interactive computer program was written to conduct the LRR calculations. Both of the treatment/solidification options were considered for a number of scenarios including type of plant (BWR or PWR) and transportation distance to the disposal site. If current trends in the escalation rates of cost components continue, the volume reduction/polyethylene solidification option will be cost effective for both BWRs and PWRs. Data indicate that a minimum net annual savings of $0.8 million per year (for a PWR shipping its waste 750 miles) and a maximum net annual savings of $9 million per year (for a BWR shipping its waste 2500 miles) can be achieved. A sensitivity analysis was performed for the burial cost escalation rate, which indicated that variation of this factor will impact the total levelized revenue requirement. The burial cost escalation rate which yields a break-even condition was determined for each scenario considered. 11 refs., 8 figs., 39 tabs

Radioactive waste processing device

International Nuclear Information System (INIS)

Ikeda, Takashi; Funabashi, Kiyomi; Chino, Koichi.

1992-01-01

In a waste processing device for solidifying, pellets formed by condensing radioactive liquid wastes generated from a nuclear power plant, by using a solidification agent, sodium chloride, sodium hydroxide or sodium nitrate is mixed upon solidification. In particular, since sodium sulfate in a resin regenerating liquid wastes absorbs water in the cement upon cement solidification, and increases the volume by expansion, there is a worry of breaking the cement solidification products. This reaction can be prevented by the addition of sodium chloride and the like. Accordingly, integrity of the solidification products can be maintained for a long period of time. (T.M.)

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

Fukazawa, Tetsuo; Ootsuka, Masaharu; Uetake, Naoto; Ozawa, Yoshihiro.

1984-01-01

Purpose: To prepare radioactive solidified wastes excellent in strength, heat resistance, weather-proof, water resistance, dampproof and low-leaching property. Method: A hardening material reactive with alkali silicates to form less soluble salts is used as a hardener for alkali silicates which are solidification filler for the radioactive wastes, and mixed with cement as a water absorbent and water to solidify the radioactive wastes. The hardening agent includes, for example, CaCO 3 , Ca(ClO 4 ) 2 , CaSiF 6 and CaSiO 3 . Further, in order to reduce the water content in the wastes and reduce the gap ratio in the solidification products, the hardener adding rate, cement adding rate and water content are selected adequately. As the result, solidification products can be prepared with no deposition of easily soluble salts to the surface thereof, with extremely low leaching of radioactive nucleides. (Kamimura, M.)

Method of disposing radioactive wastes

International Nuclear Information System (INIS)

Isozaki, Kei.

1983-01-01

Purpose : To enable safety ocean disposal of radioactive wastes by decreasing the leaching rate of radioactive nucleides, improving the quick-curing nature and increasing the durability. Method : A mixture comprising 2 - 20 parts by weight of alkali metal hydroxide and 100 parts by weight of finely powdered aqueous slags from a blast furnace is added to radioactive wastes to solidify them. In the case of medium or low level radioactive wastes, the solidification agent is added by 200 parts by weight to 100 parts by weight of the wastes and, in the case of high level wastes, the solidification agent is added in such an amount that the wastes occupy about 20% by weight in the total of the wastes and the solidification agent. Sodium hydroxide used as the alkali metal hydroxide is partially replaced with sodium carbonate, a water-reducing agent such as lignin sulfonate is added to improve the fluidity and suppress the leaching rate and the wastes are solidified in a drum can. In this way, corrosions of the vessel can be suppressed by the alkaline nature and the compression strength, heat stability and the like of the product also become excellent. (Sekiya, K.)

Radioactive gas solidification apparatus

International Nuclear Information System (INIS)

Kobayashi, Yoshihiro; Seki, Eiji; Yabu, Tomohiko; Matsunaga, Hiroyuki.

1990-01-01

Handling of a solidification container from the completion for the solidifying processing to the storage of radioactive gases by a remote control equipment such as a manipulator requires a great cost and is difficult to realize. In a radioactive gas solidification device for injection and solidification in accumulated layers of sputtered metals by glow discharge, radiation shieldings are disposed surrounding the entire container, and cooling water is supplied to a cooling vessel formed between the container and the shielding materials. The shielding materials are divided into upper and lower shielding materials, so that solidification container can be taken out from the shielding materials. As a result, the solidification container after the solidification of radioactive gases can be handled with ease. Further, after-heat can be removed effectively from the ion injection electrode upon solidifying treatment upon storage, to attain a radioactive gas solidifying processing apparatus which is safe, economical and highly reliable. (N.H.)

Method of processing radioactive liquid waste

International Nuclear Information System (INIS)

Hasegawa, Akira; Kuribayashi, Hiroshi; Soda, Kenzo; Mihara, Shigeru.

1988-01-01

Purpose: To obtain satisfactory plastic solidification products rapidly and smoothly by adding oxidizers to radioactive liquid wastes. Method: Sulfuric acid, etc. are added to radioactive liquid wastes to adjust the pH value of the liquid wastes to less than 3.0. Then, ferrous sulfates are added such that the iron concentration in the liquid wastes is 100 mg/l. Then, after adjusting pH suitably to the drying powderization by adding alkali such as hydroxide, the liquid wastes are dried and powderized. The resultant powder is subjected to plastic solidification by using polymerizable liquid unsaturated polyester resins as the solidifying agent. The thus obtained solidification products are stable in view of the physical property such as strength or water proofness, as well as stable operation is possible even for those radioactive liquid wastes in which the content ingredients are unknown. (Takahashi, M.)

Polyethylene solidification of low-level wastes

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1985-02-01

This topical report describes the results of an investigation on the solidification of low-level radioactive waste in polyethylene. Waste streams selected for this study included those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Four types of commercially available low-density polyethylenes were employed which encompass a range of processing and property characteristics. Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste and polyethylene type. Property evaluation testing was performed on laboratory-scale specimens to assess the potential behavior of actual waste forms in a disposal environment. Waste form property tests included water immersion, deformation under compressive load, thermal cycling and radionuclide leaching. Recommended waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash, and 30 wt % ion exchange resins, which are based on process control and waste form performance considerations are reported. 37 refs., 33 figs., 22 tabs

The solidification of radioactive waste

International Nuclear Information System (INIS)

Nagaya, Kiichi; Fujimoto, Yoshio; Hashimoto, Yasuo; Nomura, Ichiro

1985-01-01

A previous paper covered the decomposition and vitrification of Na 2 SO 4 (the primary component of the liquid waste from BWR) with silica. Now, in order to establish an integrated treatment system for the radioactive waste from BWR, this paper examines the effects of combining incinerator ash and other incinerator wastes with radioactive waste on the durability of the final vitrified products. A bench scale test plat consisting of a waiped file evaporator/dryer, a Joule-heated glass melter and SO 2 absorber was therefore put into operation and run safety for a period of 3000 hours. The combination of the radioactive waste with incinerator ash and the secondary waste of the incinerator was found to make no difference on the durability of the final vitrified products effecting no increase or decrease. Durability similar to that displayed in the beaker tests was proven, with the final vitrified products exhibiting a leaching rate less than 3 x 10 -4 g/cm 2 /day at 95 deg C. (author)

Development of radioactive waste treatment technique

International Nuclear Information System (INIS)

Kikuchi, Makoto; Amamiya, Shigeru; Yusa, Hideo.

1984-01-01

The techniques of radioactive waste treatment are generally reviewed, placing emphasis on volume reduction and solidification techniques. After a brief description on the general process of radioactive waste treatment, some special technologies being developed by Hitachi Ltd. are explained. From the viewpoints of the volume reduction, long term management and final disposal of wastes, the pelletization of dried waste and the solidification with inorganic substances are considered. One of the features of the pelletization system is to treat various kinds of wastes such as concentrated liquid wastes and used resins by the same system. The flow diagram of the system and its special features are shown. The volume reduction achieved by this system as compared to the conventional method is about 1/7. The first commercial plant for the treatment of concentrated liquid waste is scheduled to begin operation in June, 1984. As for the solidification technique for waste disposal, the use of cement glass is considered. The solidification system being developed is shortly described. (Aoki, K.)

Solidification of radioactive liquid wastes. A comparison of treatment options for spent resins and concentrates

International Nuclear Information System (INIS)

Roth, A.; Willmann, F.; Ebata, M.; Wendt, S.

2008-01-01

Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence of a final repository site, the built-up of additional volume has to be considered as very critical. Moreover, corrosive effects on cemented drums during long-term interim storage at the surface have raised doubts about the long-term stability of such waste products. In order to avoid such disadvantages solidification methods have been improved in order to get a well-defined product with a better load factor of wastes in the matrix. In a complete different approach, other technologies solidify the liquid radioactive wastes without adding of any inactive material by means of drying

Current high-level waste solidification technology

International Nuclear Information System (INIS)

Bonner, W.F.; Ross, W.A.

1976-01-01

Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

Method of solidifying radioactive wastes with plastics

International Nuclear Information System (INIS)

Matsuura, Hiroyuki; Yasumura, Keijiro; Minami, Yuji; Tomita, Toshihide

1980-01-01

Purpose: To prevent solidification of solidifying agents in the mixer by conducting the mixing process for the solidifying agents and the radioactive wastes at a temperature below the initiation point for the solidification of the agents thereby separating the mixing process from the solidification-integration process. Method: Catalyst such as cobalt naphthenate is charged into an unsaturated polyester resin in a mixer previously cooled, for example, to -10 0 C. They are well mixed with radioactive wastes and the mixture in the mixer is charged in a radioactive waste storage container. The temperature of the mixture, although kept at a low temperature initially, gradually increases to an ambient temperature whereby curing reaction is promoted and the reaction is completed about one day after to provide firm plastic solidification products. This can prevent the solidification of the solidifying agents in the mixer to thereby improve the circumstance's safety. (Kawakami, Y.)

Waterproofing improvement of radioactive waste asphalt solid

International Nuclear Information System (INIS)

Adachi, Katsuhiko; Yamaguchi, Takashi; Ikeoka, Akira.

1981-01-01

Purpose: To improve the waterproofing of asphalt solid by adding an alkaline earth metal salt and, further, paraffin, into radioactive liquid waste when processing asphalt solidification of the radioactive liquid waste. Method: Before processing molten asphalt solidification of radioactive liquid waste, soluble salts of alkaline earth metal such as calcium chloride, magnesium chloride, or the like is added to the radioactive liquid waste. Paraffin having a melting point of higher than 60 0 C, for example, is added to the asphalt, and waterproofing can be remarkably improved. The waste asphalt solid thus fabricated can prevent the swelling thereof, and can improve its waterproofing. (Yoshihara, H.)

Solidification of low-level wastes by inorganic binder

International Nuclear Information System (INIS)

Sasaki, M.T.; Shimojo, M.; Suzuki, K.; Kajikawa, A.; Karasawa, Y.

1995-01-01

The use of an alkali activated slag binder has been studied for solidification and stabilization of low-level wastes in nuclear power stations and spent fuel processing facilities. The activated slag effectively formed waste products having good physical properties with high waste loading for sodium sulfate, sodium nitrate, calcium pyrophosphate/phosphate and spent ion-exchange resins. Moreover, the results of the study suggest the slag has the ability to become a common inorganic binder for the solidification of various radioactive wastes. This paper also describes the fixation of radionuclides by the activated slag binder

Solidification of low-level radioactive liquid waste using a cement-silicate process

International Nuclear Information System (INIS)

Grandlund, R.W.; Hayes, J.F.

1979-01-01

Extensive use has been made of silicate and Portland cement for the solidification of industrial waste and recently this method has been successfully used to solidify a variety of low level radioactive wastes. The types of wastes processed to date include fuel fabrication sludges, power reactor waste, decontamination solution, and university laboratory waste. The cement-silicate process produces a stable solid with a minimal increase in volume and the chemicals are relatively inexpensive and readily available. The method is adaptable to either batch or continuous processing and the equipment is simple. The solid has leaching characteristics similar to or better than plain Portland cement mixtures and the leaching can be further reduced by the use of ion-exchange additives. The cement-silicate process has been used to solidify waste containing high levels of boric acid, oils, and organic solvents. The experience of handling the various types of liquid waste with a cement-silicate system is described

China's status and strategy of radioactive waste management

International Nuclear Information System (INIS)

Bi Decai

2001-01-01

China has a forty-year history of nuclear industry and nuclear technology application. Safety management of radioactive wastes has been the great concern of related regulatory authorities. After the national policy on regional disposal for low and intermediate level radioactive waste was enacted in 1992, the management of radioactive wastes gradually focused on disposal. Currently, the strategies for radioactive waste management in China are: (a) storing high level radioactive wastes temporarily and launching the study of vitrification and deep geological disposal of high level liquid waste, treating spent fuels from PWR by reprocessing; (b) implementing regional disposal policy for low and intermediate level wastes, implementing cement solidification for low and intermediate level liquid waste before disposal, carrying out bulk casting shallow land disposal technology and hydraulic-fractured cement solidification for deep geological disposal in some special regions under specific conditions, treating low and intermediate level solid radioactive wastes by cement solidification after incineration or by compressing before final disposal; (c) stabilizing the tailing repository by reinforcing embankment, constructing flood dam and overlaying plantation; and (d) developing and formulating laws, regulations, and standards to ensure safe management of radioactive wastes. When establishing standards, other than to follow the generic principles and requirements, emphasis should be placed on the following principles: safety the first, economy, disposal of radioactive wastes as focus, and introduction of international advanced standards as possible. (author)

Method of solidifying radioactive waste

International Nuclear Information System (INIS)

Hasegawa, Akira; Mihara, Shigeru; Yamashita, Koji; Sauda, Kenzo.

1988-01-01

Purpose: To obtain satisfactory plastic solidification products rapidly and more conveniently from radioactive wastes. Method: liquid wastes contain, in addition to sodium sulfate as the main ingredient, nitrates hindering the polymerizing curing reactions and various other unknown ingredients, while spent resins contain residual cationic exchange groups hindering the polymerizing reaction. Generally, as the acid value of unsaturated liquid polyester resins is lower, the number of terminal alkyd resins is small, formation of nitrates is reduced and the polymerizing curing reaction is taken place more smoothly. In view of the above, radioactive wastes obtained by dry powderization or dehydration of radioactive liquid wastes or spent resins are polymerized with unsaturated liquid polyester resins with the acid value of less than 13 to obtain plastic solidification. Thus, if the radioactive wastes contain a great amount of polymerization hindering material such as NaNO 2 , they can be solidified rapidly and conveniently with no requirement for pre-treatment. (Kamimura, Y.)

MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES

International Nuclear Information System (INIS)

Solbrig, Charles W.; Bateman, Kenneth J.

2010-01-01

The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesn't cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, ''the length deficit,'' produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

Solidification of radioactive liquid wastes, Treatment options for spent resins and concentrates - 16405

International Nuclear Information System (INIS)

Roth, Andreas

2009-01-01

Ion exchange is one of the most common and effective treatment methods for liquid radioactive waste. However, spent ion exchange resins are considered to be problematic waste that in many cases require special approaches and pre-conditioning during its immobilization to meet the acceptance criteria for disposal. Because of the function that they fulfill, spent ion exchange resins often contain high concentrations of radioactivity and pose special handling and treatment problems. Another very common method of liquid radioactive waste treatment and water cleaning is the evaporation or diaphragm filtration. Both treatment options offer a high volume reduction of the total volume of liquids treated but generate concentrates which contain high concentrations of radioactivity. Both mentioned waste streams, spent resins as well as concentrates, resulting from first step liquid radioactive waste treatment systems have to be conditioned in a suitable manner to achieve stable waste products for final disposal. Spent resin and concentrate treatment often appear as a specific task in decommissioning projects, because in the past those waste streams typically had been stored in tanks for the lifetime of the plant and needs to be retrieved, conditioned and packed prior to dismantling activities. Additionally a large amount of contaminated liquids will be generated by utilizing decontamination processes and needs to be processed further on. Such treatment options need to achieve waste products acceptable for final disposal, because due to the closure of the site no interim storage can be envisaged. The most common method of treatment of such waste streams is the solidification in a solid matrix with additional inactive material like cement, polymer etc. In the past good results have been achieved and the high concentration of radioactivity can be reduced by adding the inactive material. On the other hand, under the environment of limited space for interim storage and the absence

The influence of radioactive waste solidification methods and storage conditions on the migration and dispersion of radionuclides in the terrestrial environment

International Nuclear Information System (INIS)

Golinski, M.; Ferenc, M.; Tomczak, W.; Cholerzynski, A.

1979-01-01

In the first part of the paper a survey of literature data on the methods and apparatus used for solidification of low- and intermediate-level radioactive wastes is done and the methods for the determination of the solidification product properties are discussed. The second part of the paper contains the experimental leachability data for 60 Co, 90 Sr and 137 Cs from simulated radioactive waste solidification products obtained with the help of bitumen, cement and urea formaldehyde resin. The leachability of radionuclides from the bituminization products decreases in the following order: 137 Cs 90 Sr 60 Co while that form concrete and the urea formaldehyde resin blocks as follows: 137 Cs 60 Co 90 Sr. A very good resistance of bitumen blocks against changeable atmospheric conditions and saline has been observed. The results obtained show that bitumen is the best binder while urea formaldehyde resin is the worst. (author)

Method of solidifying radioactive solid wastes

International Nuclear Information System (INIS)

Fukazawa, Tetsuo; Kawamura, Fumio; Kikuchi, Makoto.

1984-01-01

Purpose: To obtain solidification products of radioactive wastes satisfactorily and safely with no destruction even under a high pressure atmosphere by preventing the stress concentration by considering the relationships of the elastic module between the solidifying material and radioactive solid wastes. Method: Solidification products of radioactive wastes with safety and securing an aimed safety ratio are produced by conditioning the modules of elasticity of the solidifying material equal to or less than that of the radioactive wastes in a case where the elastic module of radioactive solid wastes to be solidified is smaller than that of the solidifying material (the elastic module of wastes having the minimum elastic module among various wastes). The method of decreasing the elastic module of the solidifying material usable herein includes the use of such a resin having a long distance between cross-linking points of a polymer in the case of plastic solidifying materials, and addition of rubber-like binders in the case of cement or like other inorganic solidifying materials. (Yoshihara, H.)

Radioactive gas solidification treatment device

International Nuclear Information System (INIS)

Igarashi, Ryokichi; Watanabe, Yu; Seki, Eiji.

1992-01-01

In a radioactive gas solidification treatment device by using sputtering, spiral pipelines are disposed with a gap therebetween for cooling an ion injection electrode by passing cooling water during operation of the solidification treatment. During the operation of the solidification treatment, cooling water is passed in the pipelines to cool the ion injection electrode. During storage, a solidification vessel is cooled by natural heat dissipation from an exposed portion at the surface of the solidification vessel. Accordingly, after-heat of radioactive gas solidified in a metal accumulation layer can be removed efficiently, safely and economically to improve the reliability. (N.H.)

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

Maeda, Masahiko; Kira, Satoshi; Watanabe, Naotoshi; Nagaoka, Takeshi; Akane, Junta.

1982-01-01

Purpose: To obtain solidification products of radioactive wastes having sufficient monoaxial compression strength and excellent in water durability upon ocean disposal of the wastes. Method: Solidification products having sufficient strength and filled with a great amount of radioactive wastes are obtained by filling and solidifying 100 parts by weight of chlorinated polyethylene resin and 100 - 500 parts by weight of particular or powderous spent ion exchange resin as radioactive wastes. The chlorinated polyethylene resin preferably used herein is prepared by chlorinating powderous or particulate polyethylene resin in an aqueous suspending medium or by chlorinating polyethylene resin dissolved in an organic solvent capable of dissolving the polyethylene resin, and it is crystalline or non-crystalline chlorinated polyethylene resin comprising 20 - 50% by weight of chlorine, non-crystalline resin with 25 - 40% by weight of chlorine being particularly preferred. (Horiuchi, T.)

A comparison of solidification media for the stabilization of low- level radioactive wastes

International Nuclear Information System (INIS)

Cowgill, M.G.

1991-10-01

When requirements exist to stabilize low-level radioactive waste (LLW) prior to disposal, efforts to achieve this stability often center on the mixing of the waste with a solidification medium. Although historically the medium of choice has been based on the use of portland cement as the binder material, several other options have been developed and subsequently implemented. These include thermoplastic polymers, thermosetting polymers and gypsum. No one medium has thus far been successful in providing stability to all forms of LLW. The characteristics and attributes of these different binder materials are reviewed and compared. The aspects examined include availability of information, limitations to use, sensitivity to process or waste chemistry changes, radionuclide retention ability, modeling of radionuclide release processes, ease and safety of use, and relative costs

The management of high-level radioactive wastes

International Nuclear Information System (INIS)

Lennemann, Wm.L.

1979-01-01

The definition of high-level radioactive wastes is given. The following aspects of high-level radioactive wastes' management are discussed: fuel reprocessing and high-level waste; storage of high-level liquid waste; solidification of high-level waste; interim storage of solidified high-level waste; disposal of high-level waste; disposal of irradiated fuel elements as a waste

Processing and solidification of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Kelley, J.A.

1981-01-01

The entire flowsheet for processing and solidification of Savannah River Plant (SRP) high-level wastes has been demonstrated. A new small-scale integrated pilot plant is operating with actual radioactive wastes, and large-scale equipment is being demonstrated with nonradioactive simulated wastes. Design of a full-scale waste solidification plant is in progress. Plant construction is expected to begin in 1983, and startup is anticipated in 1988. The plant will poduce about 500 cans of glass per year with each can containing about 1.5 tons of glass

Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices

International Nuclear Information System (INIS)

Donato, A.; Arcuri, L.; Dotti, M.; Pace, A.; Pietrelli, L.; Ricci, G.; Basta, M.; Cali, V.; Pagliai, V.

1989-05-01

This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

Glass-solidification method for high level radioactive liquid waste

International Nuclear Information System (INIS)

Kawamura, Kazuhiro; Kometani, Masayuki; Sasage, Ken-ichi.

1996-01-01

High level liquid wastes are removed with precipitates mainly comprising Mo and Zr, thereafter, the high level liquid wastes are mixed with a glass raw material comprising a composition having a B 2 O 3 /SiO 2 ratio of not less than 0.41, a ZnO/Li 2 O ratio of not less than 1.00, and an Al 2 O 3 /Li 2 O ratio of not less than 2.58, and they are melted and solidified into glass-solidification products. The liquid waste content in the glass-solidification products can be increased up to about 45% by using the glass raw material having such a predetermined composition. In addition, deposition of a yellow phase does not occur, and a leaching rate identical with that in a conventional case can be maintained. (T.M.)

Solidification of low-level waste - a dilemma for the small user

International Nuclear Information System (INIS)

Harris, S.; Gilmore, A.

1980-01-01

The requirement that radioactive waste for sea disposal must be solidified by the originator is discussed. Attempts to solidify small quantities of radioactive waste such as contaminated oils and labelled benzyopyrene with other solvents are described. Encapsulation media tested were concrete and interior and exterior grade Polyfilla (a plaster and cellulose based filler). Problems were presented by the difficulty of mixing the materials and by the maximum uptake of solvents while still allowing solidification. In all cases a soft crumbling material resulted. It is concluded that solidification processing on a small scale does not make economic or scientific sense and that if solidification is necessary it would be better carried out as a national operation by collecting liquids from users. (U.K.)

System for disposing of radioactive waste

International Nuclear Information System (INIS)

Gablin, K.A.; Hansen, L.J.

1979-01-01

A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container. 30 claims

System for disposing of radioactive waste

International Nuclear Information System (INIS)

Gablin, K.A.; Hansen, L.J.

1977-01-01

A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container

Modified sulfur cement solidification of low-level wastes

Energy Technology Data Exchange (ETDEWEB)

1985-10-01

This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended.

Modified sulfur cement solidification of low-level wastes

International Nuclear Information System (INIS)

1985-10-01

This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended

Mixed and chelated waste test programs with bitumen solidification

International Nuclear Information System (INIS)

Simpson, S.I.; Morris, M.; Vidal, H.

1988-01-01

This paper presents the results of bitumen solidification tests on mixed wastes and chelated wastes. The French Atomic Energy Commission (CEA) performed demonstration tests on radioactive wastes contaminated with chelating agents for Associated Technologies, Inc. (ATI). The chelated wastes were produced and concentrated by Commonwealth Edison Co. as a result of reactor decontamination at Dresden Nuclear Station, Unit 1. Law Engineering in Charlotte, N. C. produced samples and performed tests on simulated heavy metal laden radioactive waste (mixed) to demonstrate the quality of the bituminous product. The simulation is intended to represent waste produced at Oak Ridge National Labs operated by Martin-Marietta

Development of stable solidification methods for toxic lead oxide in radioactive wastes

International Nuclear Information System (INIS)

Hitoshi Mimura; Shingo Ikeda; Yuichi Niibori

2009-01-01

The objective of this study is to develop the advanced solidification methods for the toxic lead oxide contained in radioactive wastes and to examine their chemical durability in terms of leachability and surface alteration; the solidification characteristics and leachability for the following three kinds of solidified products immobilizing lead were examined, and the experimental results were summarized as follows. (a) Mineral solidified products: A-zeolite or fly ash (FA) was used as a binder, and NaAlO 2 and Na 2 SiO 3 were mixed as additives. The leachability of lead ions in pure water was considerably lowered by the heat treatment at higher temperature (1,000 degree C), and the concentration of lead ions leached was under criterion value of 0.3 mg/l. The products prepared by mixing A-zeolite and fly ash also had low leachability under 0.3 mg/l even in the saturated Ca(OH)2 solution. (b) Melted solidified products: A-zeolite or fly ash was used as a binder and glass-forming reagents of B 2 O 3 and NaH 2 PO 4 were used as additives. The XRD peaks assigned PbO were not observed in all products. The products for the mixtures of FA:NaH 2 PO 4 :PbO (2:2:1 and 3:1:1) had low leachability under criterion value in both leachants of deionized water and saturated Ca(OH) 2 solution. (c) Phosphate ceramics products: the chemically bonded phosphate ceramics were produced by using MgKPO 4 , MgHPO 4 , Zr(HPO 4 ) 2 , potassium iron phosphate and sodium iron phosphate, and FA was used as additives. In particular, by using MgHPO 4 , the leachability of the products was lowered less than 0.3 mg/l in both leachants. The phosphate ceramics products and melted solidified products are favorable as the waste solid forms immobilizing lead. In particular, novel ceramics products have advantages in the simple solidification procedure similarly to the cement products. As for mineral solidification, natural zeolites and FA as binder also useful from the viewpoint of cost efficiency

Method of reprocessing radioactive asphalt solidification products

International Nuclear Information System (INIS)

Nakaya, Iwao; Murakami, Tadashi; Miyake, Takafumi; Inagaki, Yuzo.

1986-01-01

Purpose: To obtain heat-stable solidification products and decrease the total volume thereof by modifying the solidified form by the reprocessing of existent radioactive asphalt solidification products. Method: Radioactive asphalt solidification products are heated into a fluidized state. Then, incombustible solvents such as perchloroethylene or trichloroethylene are added to a dissolving tank to gradually dissolve the radioactive asphalt solidification products. Thus, organic materials such as asphalts are transferred into the solvent layer, while inorganic materials containing radioactive materials remain as they are in the separation tank. Then, the inorganic materials containing the radioactive materials are taken out and then solidified, for example, by converting them into a rock or glass form. (Kawakami, Y.)

Solidification processing method for radioactive waste

International Nuclear Information System (INIS)

Hiraki, Akimitsu; Tanaka, Keiji; Heta, Katsutoshi.

1991-01-01

The pressure in a vessel containing radioactive wastes is previously reduced and cement mortar prepared by kneading cement, sand and kneading agent with water is poured under shaking substantially to the upper end of the vessel. After the lowering of the mortar level due to the deforming has been terminated, the pressure is increased gradually. Then, the cement mortar is further poured substantially to the upper end of the vessel again. With such a two step pouring method, spaces other than the radioactive wastes in the vessel can be filled substantially completely with the cement mortar. Accordingly, it is possible to avoid the problem in view of the strength due to the formation of gaps at the inside of the vessel, or leaching of radioactive materials due to the intrusion of water into the gaps. Further, if washing water is reutilized as water for kneading or washing after the precipitation of the solid contents, the amount of the secondary wastes generated can be reduced. (T.M.)

Radioactive waste management

International Nuclear Information System (INIS)

Blomek, D.

1980-01-01

The prospects of nuclear power development in the USA up to 2000 and the problems of the fuel cycle high-level radioactive waste processing and storage are considered. The problems of liquid and solidified radioactive waste transportation and their disposal in salt deposits and other geologic formations are discussed. It is pointed out that the main part of the high-level radioactive wastes are produced at spent fuel reprocessing plants in the form of complex aqueous mixtures. These mixtures contain the decay products of about 35 isotopes which are the nuclear fuel fission products, about 18 actinides and their daughter products as well as corrosion products of fuel cans and structural materials and chemical reagents added in the process of fuel reprocessing. The high-level radioactive waste management includes the liquid waste cooling which is necessary for the short and middle living isotope decay, separation of some most dangerous components from the waste mixture, waste solidification, their storage and disposal. The conclusion is drawn that the seccessful solution of the high-level radioactive waste management problem will permit to solve the problem of the fuel cycle radioactive waste management as a whole. The salt deposits, shales and clays are the most suitable for radioactive waste disposal [ru

Management of radioactive wastes in China

International Nuclear Information System (INIS)

Pan Ziqiang

1994-01-01

The policy and principles on management of radioactive wastes are stipulated. Cement solidification and bituminization unit has come into trial run. Solid radioactive waste is stored in tentative storage vault built in each of nuclear facilities. Seventeen storages associated with applications of nuclear technology and radioisotopes have been built for provinces. Disposal of low and intermediate level radioactive wastes pursues the policy of 'regional disposal'. Four repositories have been planned to be built in northwest, southwest, south and east China respectively. A program for treatment and disposal of high level radioactive waste has been made

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro's Bruce Nuclear Generating Station 'A'

International Nuclear Information System (INIS)

Day, J.E.; Baker, R.L.

1995-01-01

Ontario Hydro at the Bruce Nuclear Generating Station 'A' has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

Sponsored research on radioactive waste management

International Nuclear Information System (INIS)

1983-01-01

The report is in chapters entitled: introduction (background, responsibilities, options, structure of the programme); strategy development; disposal of accumulations; disposal of radioactive waste arisings; quality assurance for waste conditioning quality assurance related to radioactive waste disposal (effectiveness of different rock types as natural barriers to the movement of radioactivity, and non-site specific factors in the design of repositories; radiological assessment; environmental studies; research and development to meet requirements specific to UKAEA wastes; long term research (processes for the solidification of highly active liquid wastes); plutonium contamination waste minimisation. (U.K.)

Evolution in radioactive waste countermeasures

International Nuclear Information System (INIS)

Moriguchi, Yasutaka

1984-01-01

The establishment of radioactive waste management measures is important to proceed further with nuclear power development. While the storage facility projects by utilities are in progress, large quantity of low level wastes are expected to arise in the future due to the decommissioning of nuclear reactors, etc. An interim report made by the committee on radioactive waste countermeasures to the Atomic Energy Commission is described as follows: the land disposal measures of ultra-low level and low level radioactive wastes, that is, the concept of level partitioning, waste management, the possible practice of handling wastes, etc.; the treatment and disposal measures of high level radioactive wastes and transuranium wastes, including task sharing among respective research institutions, the solidification/storage and the geological formation disposal of high level wastes, etc. (Mori, K.)

The solidification of low level radioactive organic fluids with Envirostone Gypsum Cement

International Nuclear Information System (INIS)

Rosenstiel, T.L.; Lange, R.G.

1984-01-01

The primary method for the management of low level radioactive waste (LLW) has been and continues to be the isolation of the waste in a solid mass. Of the four typical LLW streams, organic fluids pose the most significant waste isolation problem. The organic fluids comprised of lubrication oils, hydraulic fluids, sludges, scintillation fluids, etc., result from the operation and maintenance of nuclear power generating stations, research activities, tooling operations, and diagnostic analyses. The United States Gypsum Company developed the patented Envirostone Gypsum Cement system for the solidification of all types of low level radioactive wastes to facilitate handling and transportation to regulated LLW disposal sites. For the solidification of organic fluids, Envirostone Gypsum Cement is used in conjunction with Envirostone Emulsifier, selected for its ability to emulsify a broad range of organic fluids in aqueous solutions. In the solidification process it is theorized that as the crystalline matrix of the gypsum forms, the micelles of the emulsifier behave as a chemical bridge which draws the organic fluid into the crystalline structure via the hydration water. Initial testing of physical properties of solidified waste forms, including leachability, per the requirements and the procedures specified for 10 CFR Part 61 as outlined in the Branch Technical Position Report from the United States Nuclear Regulatory Commission were in progress as of the writing of this paper. Upon completion of this testing a Topical Report will be submitted to the USNRC for review and approval. The presentation reviews field experience in the use of Envirostone Gypsum Cement for the solidification of low level radioactive organic fluids from nuclear power generating stations and makes an economic comparison between Envirostone Gypsum Cement and portland cement systems

INEL studies concerning solidification of low-level waste in cement

International Nuclear Information System (INIS)

Mandler, J.W.

1989-01-01

The Idaho National Engineering Laboratory (INEL) has performed numerous studies addressing issues concerning the solidification of low-level radioactive waste in cement. These studies have been performed for both the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE). This short presentation will only outline the major topics addressed in some of these studies, present a few conclusions, and identify some of the technical concerns we have. More details of the work and pertinent results will be given in the Working Group sessions. The topics that have been addressed at the INEL which are relevant to this Workshop include (1) solidification of ion-exchange resins and evaporator waste in cement at commercial nuclear power plants, (2) leachability and compressive strength of power plant waste solidified in cement, (3) suggested guidelines for preparation of a solid waste process control program (PCP), (4) cement solidification of EPICOR-II resin wastes, and (5) performance testing of cement-solidified EPICOR-II resin wastes

MANAGEMENT OF RADIOACTIVE WASTES IN CHINA

Institute of Scientific and Technical Information of China (English)

潘自强

1994-01-01

The policy and principles on management of radioactive wastes are stipulated.Cement solidification and bituminization unit has come into trial run.Solid radioactive waste is stored in tentative storage vault built in each of nuclear facilities.Seventeen storages associated with applications of nuclear technology and radioisotopes have been built for provinces.Disposal of low and intermediate level radioactive wastes pursues the policy of “regional disposal”.Four repositories have been planned to be built in northwest.southwest,south and east China respectively.A program for treatment and disposal of high level radioactive waste has been made.

Method and apparatus for solidifying radioactive waste

International Nuclear Information System (INIS)

Kadota, Hiroko; Kikuchi, Makoto; Tsuchiya, Hiroyuki; Tamada, Shin.

1989-01-01

The present invention concerns a method of solidifying radioactive wastes that generate heat with water curing solidifying material and the object there of is suppress the effect of heat generation of the wastes given on the solidification material. That is, it is a feature of the invention to inject water content contained in the water curable solidification material in the form of ice into the wastes. Thus, since the water content in the water curable solidification material is ice, the solidification products can be obtained by way of the following three steps: (1) ice is dissolved into water, (2) solid content of the solidification material is dissolved into water, and(3) curing reaction of the solidification material is started. Acccordingly, since the heat generated from the wastes contributes as heat of reaction when ice is dissolved into water till the solidification material has been completely filled, promotion for the curing reaction causing problems so far can be suppressed to enable easy filling. Then, after the completion of the filling of the solidification material, the heat of the wastes has an effect of promoting the second and the third steps described above to accelerate the curing reaction. (K.M.)

Method of processing liquid wastes containing radioactive materials

International Nuclear Information System (INIS)

Matsumoto, Kaname; Shirai, Takamori; Nemoto, Kuniyoshi; Yoshikawa, Jun; Matsuda, Takeshi.

1983-01-01

Purpose: To reduce the number of solidification products by removing, particularly, Co-60 that is difficult to remove in a radioactive liquid wastes containing a water-soluble chelating agent, by adsorbing Co-60 to a specific chelating agent. Method: Liquid wastes containing radioactive cobalt and water-soluble chelating agent are passed through the layer of less water-soluble chelating agent that forms a complex compound with cobalt in an acidic pH region. Thus, the chelating compound of radioactive cobalt (particularly Co-60) is eliminated by adsorbing the same on a specific chelating agent layer. The chelating agent having Co-60 adsorbed thereon is discarded as it is through the cement- or asphalt-solidification process, whereby the number of solidification products to be generated can significantly be suppressed. (Moriyama, K.)

Solidification of hazardous and mixed radioactive waste at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Boehmer, A.M.; Larsen, M.M.

1986-01-01

EG and G Idaho has initiated a program to develop treatment options for the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory (INEL). This program includes development of solidification methods for some of these wastes. Testing has shown that toxic wastes can be successfully solidified using cement, cement-silicate, or ENVIROSTONE binders to produce nontoxic stable waste forms for safe, long term disposal. This paper presents the results of the solidification development program conducted at the INEL by EG and G Idaho

Solidification of hazardous and mixed radioactive waste at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Boehmer, A.M.; Larsen, M.M.

1986-03-01

EG and G Idaho has initiated a program to develop treatment options for the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory (INEL). This program includes development of solidification methods for some of these wastes. Testing has shown that toxic wastes can be successfully solidified using cement, cement-silicate, or ENVIROSTONE binders to produce nontoxic stable waste forms for safe, long term disposal. This paper presents the results of the solidification development program conducted at the INEL by EG and G Idaho

Radioactive waste processing method and processing device therefor

International Nuclear Information System (INIS)

Matsuo, Toshiaki; Nishi, Takashi; Noge, Kenji; Matsuda, Masami; Takeshi, Kiyotaka

1998-01-01

Each predetermined amount of aggregates such as cements and sands as water-hardening solidification materials and kneading water are charged from a solidification material containing vessel, an aggregate containing vessel and a kneading water containing vessel to a kneading vessel of a paste supply device. The cements, the sands and the kneading water are kneaded by the rotation of a kneader. A produced solidification material paste is charged from the kneader to a drum through a paste transporting pump. Miscellaneous radioactive solid wastes have been filled in a drum. The solidification paste produced while supplying the cements, the sands and the kneading water into the kneader is discharged from the kneader. Since increase of viscosity of the solid material paste in the kneader is suppressed, the solidification paste can be easily flown into narrow gaps between radioactive miscellaneous solid wastes in the drum. (I.N.)

Radioactive wastes and their disposal

International Nuclear Information System (INIS)

Neumann, L.

1984-01-01

The classification of radioactive wastes is given and the achievements evaluated in the disposal of radioactive wastes from nuclear power plants. An experimental pilot unit was installed at the Jaslovske Bohunice nuclear power plant for the bituminization of liquid radioactive wastes. UJV has developed a mobile automated high-output unit for cementation. In 1985 the unit will be tested at the Jaslovske Bohunice and the Dukovany nuclear power plants. A prototype press for processing solid wastes was manufactured which is in operation at the Jaslovske Bohunice plant. A solidification process for atypical wastes from long-term storage of spent fuel elements has been developed to be used for the period of nuclear power plant decommissioning. (E.S.)

Microwave solidification development for Rocky Flats waste

Energy Technology Data Exchange (ETDEWEB)

Dixon, D.; Erle, R.; Eschen, V. [and others

1994-04-01

The Microwave Engineering Team at the Rocky Flats Plant has developed a production-scale system for the treatment of hazardous, radioactive, and mixed wastes using microwave energy. The system produces a vitreous final form which meets the acceptance criteria for shipment and disposal. The technology also has potential for application on various other waste streams from the public and private sectors. Technology transfer opportunities are being identified and pursued for commercialization of the microwave solidification technology.

Microwave solidification development for Rocky Flats waste

International Nuclear Information System (INIS)

Dixon, D.; Erle, R.; Eschen, V.

1994-04-01

The Microwave Engineering Team at the Rocky Flats Plant has developed a production-scale system for the treatment of hazardous, radioactive, and mixed wastes using microwave energy. The system produces a vitreous final form which meets the acceptance criteria for shipment and disposal. The technology also has potential for application on various other waste streams from the public and private sectors. Technology transfer opportunities are being identified and pursued for commercialization of the microwave solidification technology

Solidification of intermediate level liquid waste - ILLW, CEMEX waste form qualification

International Nuclear Information System (INIS)

D'Andrea, V.; Guerra, M.; Pancotti, F.; Maio, V.

2015-01-01

In the Sogin EUREX Facility about 125 m 3 of intermediate level radioactive waste and about 113 m 3 of low level radioactive waste, produced during the re-processing of MTR and CANDU fuel, are stored. Solidification of these wastes is planned in order to fulfill the specific requirements established by the Safety Authority, taking into account the criteria set up in a Technical Guide on the issue of radioactive waste management. The design of a cementation plant (CEMEX) of all liquid radioactive wastes is currently ongoing. The process requires that the liquid waste is neutralized with NaOH (NaOH 19 M) and metered into 440 liter drum together with the cement, while the mixture is stirred by a lost paddle ('in drum mixing process'). The qualification of the Waste Form consists of all the activities demonstrating that the final cemented product has the minimum requirements (mechanical, chemical and physical characteristics) compliant with all the subsequent management phases: long-term interim storage, transport and long-term disposal of the waste. All tests performed to qualify the conditioning process for immobilizing first extraction cycle (MTR and CANDU) and second extraction cycle liquid wastes, gave results in compliance with the minimum requirements established for disposal

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario hydro's Bruce nuclear generating station open-quotes Aclose quotes

International Nuclear Information System (INIS)

Day, J.E.; Baker, R.L.

1994-01-01

Ontario Hydro at the Bruce Nuclear Generating Station open-quotes Aclose quotes has undertaken a program to render the station's liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology

Solidification of commercial and defense low-level radioactive waste in polyethylene

International Nuclear Information System (INIS)

Franz, E.M.; Heiser, L.H.; Colombo, P.

1987-08-01

A process was developed for the solidification of salt wastes, incinerator ash and ion-exchange resins in polyethylene. Of the salt wastes, sodium sulfate and boric acid are representative of the wastes produced at commercial nuclear facilities while sodium nitrate in a typical high-volume waste generated at defense-related facilities. Ease of processibility and high loading efficiencies were obtained through the use of low-density polyethylene with melt indices ranging from 2.0 to 55.0 g/minute. The process utilized a commercially available single-screw extruder to incorporate the wastes into the polyethylene at about 120 0 C to produce a homogeneous mixture. Although present studies utilize dry wastes, wet wastes can also be processed using vented extruders of the type used commercially for the bitumen solidification process. Tests were performed on the waste forms to determine leachability and mechanical properties. To confirm the compatibility of polyethylene and nitrate salt waste at elevated temperatures, the self-ignition temperatures were measured and a differential scanning calorimeter was used to characterize the thermal behavior of oxidizing compounds contained in the simulated waste, as well as the real Savannah River Plant waste. No exothermic reactions were observed over the temperature range studied from 50 0 C to 400 0 C. 18 refs., 7 figs., 8 tabs

Hazardous and mixed waste solidification development conducted at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Boehmer, A.M.; Larsen, M.M.

1986-04-01

EG and G Idaho, Inc., has initiated a program to develop safe, efficient, cost-effective solidification treatment methods for the disposal of some of the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory (INEL). Testing has shown that Extraction Procedure (EP) toxic wastes can be successfully solidified using cement, cement-silicate, or ENVIROSTONE binders to produce nontoxic stable waste forms for safe, long-term disposal as general or low-level waste, depending upon the radioactivity. The results of the solidification development program are presented in this report

Safe operation of existing radioactive waste management facilities at Dalat Nuclear Research Institute

International Nuclear Information System (INIS)

Pham Van Lam; Ong Van Ngoc; Nguyen Thi Nang

2000-01-01

The Dalat Nuclear Research Reactor was reconstructed from the former TRIGA MARK-II in 1982 and put into operation in March 1984. The combined technology for radioactive waste management was newly designed and put into operation in 1984. The system for radioactive waste management at the Dalat Nuclear Research Institute (DNRI) consists of radioactive liquid waste treatment station and disposal facilities. The treatment methods used for radioactive liquid waste are coagulation and precipitation, mechanical filtering and ion- exchange. Near-surface disposal of radioactive wastes is practiced at DNRI In the disposal facilities eight concrete pits are constructed for solidification and disposal of low level radioactive waste. Many types of waste generated in DNRI and in some Nuclear Medicine Departments in the South of Vietnam are stored in the disposal facilities. The solidification of sludge has been done by cementation. Hydraulic compactor has done volume reduction of compatible waste. This paper presents fifteen-years of safe operation of radioactive waste management facilities at DNRI. (author)

Comparison of costs for solidification of high-level radioactive waste solutions: glass monoliths vs metal matrices

International Nuclear Information System (INIS)

Jardine, L.J.; Carlton, R.E.; Steindler, M.J.

1981-05-01

A comparative economic analysis was made of four solidification processes for liquid high-level radioactive waste. Two processes produced borosilicate glass monoliths and two others produced metal matrix composites of lead and borosilicate glass beads and lead and supercalcine pellets. Within the uncertainties of the cost (1979 dollars) estimates, the cost of the four processes was about the same, with the major cost component being the cost of the primary building structure. Equipment costs and operating and maintenance costs formed only a small portion of the building structure costs for all processes

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

Tomita, Toshihide; Minami, Yuji; Matsuura, Hiroyuki; Kageyama, Hisashi; Kobori, Junzo.

1984-01-01

Purpose: To perform the curing sufficiently even when copper hydroxide that interferes the curing reaction is contained in radioactive wastes. Method: Solidification of radioactive wastes containing copper hydroxide using thermoset resins is carried out under the presence of an alkaline material. The thermoset resin used herein is an polyester resin comprising unsaturated polyester and a polymerizable monomer. The alkaline substance usable herein can include powder or an aqueous solution of hydroxides or oxides of sodium, magnesium, calcium or the like. (Yoshino, Y.)

Alternative solidification techniques for radioactive ion exchange resins and liquid concentrates

International Nuclear Information System (INIS)

Thegerstroem, C.

1980-01-01

Methods, that are used or are under development for solidification of radioactive ion exchange resins or liquid concentrates, utilize normally cement, bitumen or some polymere as matrix material. This report contains a review and a description of these solidification processes and their products, especially of relatively new techniques that are under development in different countries. It is possible that solidification in thermosetting resins will be more used in the future, especially when product quality requirements are high (for instance when solidifying medium level resins) or when special waste categories has to be solidified. However it is not probable that thermosetting resins will be extensively used in a broad application as matrix material. In that case the methods are to complicated and expensive compared to, for instance, solidification in concrete. Systems for incorporation in polyesteremulsions (Dow-process) have a potential as they are quite simple and can accept a large variation of liquid wastes. Some methods in an early stage of development (for instance Inert Carrier Radwaste Process) will have to be tested in active application before they can be further evaluated. (author)

Management of radioactive waste nuclear power plants

International Nuclear Information System (INIS)

Dlouhy, Z.; Marek, J.

1976-01-01

The authors give a survey of the sources, types and amounts of radioactive waste in LWR nuclear power stations (1,300 MWe). The amount of solid waste produced by a Novovorenezh-type PWR reactor (2 x 400 resp. 1 x 1,000 MWe) is given in a table. Treatment, solidification and final storage of radioactive waste are shortly discussed with special reference to the problems of final storage in the CSR. (HR) [de

Radioactive waste management in West Germany

Energy Technology Data Exchange (ETDEWEB)

Krause, H [Kernforschungszentrum Karlsruhe G.m.b.H. (Germany, F.R.)

1978-01-01

The technologies developed in West Germany for radioactive waste management are widely reviewed. The first topic in this review paper is the disposal of low- and middle-level radioactive liquid wastes. Almost all these liquid wastes are evaporated, and the typical decontamination factor attained is 10/sup 4/ -- 10/sup 6/. The second topic is the solidification of residuals. Short explanation is given to bituminization and some new processes. The third topic is high-level liquid wastes. Degradation of glass quality due to various radiation is discussed. Embedding of small glass particles containing radioactive wastes into metal is also explained. Disposals of low-level solid wastes and the special wastes produced from reprocessing and mixed oxide fuel fabrication are explained. Final disposal of radioactive wastes in halite is discussed as the last topic. Many photographs are used to illustrate the industrial or experimental use of those management methods.

Process control of Low and Intermediate-level radioactive wastes solidification

International Nuclear Information System (INIS)

1993-01-01

Safety guidelines issued by the Spanish Council of Nuclear Safety (CSN) with basic criteria which must be adopted for the control of the Process for wastes solidification, establishing, in addition, a series of protocols and basic contents to assist the elaboration of Process Control Programs

Evaluation of process alternatives for solidification of the West Valley high-level liquid wastes

International Nuclear Information System (INIS)

Holton, L.K.; Larson, D.E.

1982-01-01

The Department of Energy (DOE) established the West Valley Solidification Project (WVSP) in 1980. The project purpose is to demonstrate removal and solidification of the high-level liquid wastes (HLLW) presently stored in tanks at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York. As part of this effort, the Pacific Northwest Laboratory (PNL) conducted a study to evaluate process alternatives for solidifcation of the WNYNSC wastes. Two process approaches for waste handling before solidification, together with solidification processes for four terminal and four interim waste forms, were considered. The first waste-handling approach, designated the salt/sludge separation process, involves separating the bulk of the nonradioactive nuclear waste constituents from the radioactive waste constituents, and the second waste-handling approach, designated the combined-waste process, involves no waste segregation prior to solidification. The processes were evaluated on the bases of their (1) readiness for plant startup by 1987, (2) relative technical merits, and (3) process cost. The study has shown that, based on these criteria, the salt/sludge separation process with a borosilicate glass waste form is preferred when producing a terminal waste form. It was also concluded that if an interim waste form is to be used, the preferred approach would be the combined waste process with a fused-salt waste form

Safety disposal studies of radioactive and hazardous wastes using cement

International Nuclear Information System (INIS)

Aly, M.M.E.

2000-01-01

radioactive waste is generated from the production of nuclear energy and from the use of radioactive materials applications, agriculture and medicine. the important of safe management of radioactive waste for the protection of human health and the environment has long been recognized. conditioning of radioactive waste is the transform of radioactive waste into a suitable form for storage and disposal. common immobilization methods include solidification of low radioactive waste in cement or bitumen.in order to improve cement properties to decrease the release of liquid radioactive waste into the environment and its dispersion to a level where the risks to individuals, population and the environment

Method and apparatus for glass solidification porcessing for radioactive liquid waste

International Nuclear Information System (INIS)

Torada, Shin-ichiro; Masaki, Toshio; Sakai, Akira.

1989-01-01

Glass material supplied to a glass melting furnace is made in the form of a glass container. Then, radioactive liquid wastes are directly injected into the glass vessel and the glass vessel injected with the radioactive liquid wastes is charged into the glass melting furnace. The glass material and the radioactive liquid wastes are supplied simultaneously to the glass melting furnace. Then, corresponding to the amount of the glass material used for the glass vessel, the amount of the radioactive liquid wastes injected to the inside thereof is controlled to thereby set the mixing ratio between the glass material and the radioactive liquid wastes. Further, by controlling the number of the glass vessels injected with the radioactive liquid wastes to be charged into the glass melting furnace, the amount of supplying the radioactive liquid wastes and the glass material is controlled. This can easily maintain constant the amount of the glass material and the radioacative liquid wastes supplied to the glass melting furnace and the mixing ratio thereof. (T.M.)

Volume reduction and solidification of radioactive waste incineration ash with waste glass

International Nuclear Information System (INIS)

Koyama, Hidemi; Kobayashi, Masayuki

2007-01-01

The low-level radioactive waste generated from research institutions and hospitals etc. is packed into a container and is kept. The volume reduced state or the unprocessed state by incineration or compression processing are used because neither landfill sites nor disposal methods have been fixed. Especially, because the bulk density is low, and it is easy to disperse, the low-level radioactive waste incineration ash incinerated for the volume reduction is a big issue in security, safety, stability in the inventory location. A safe and appropriate disposal processing method is desired. When the low temperature sintering method in the use of the glass bottle cullet was examined, volume reduction and stabilization of low-level radioactive waste incineration ash were verified. The proposed method is useful for the easy treatment of the low-level radioactive waste incineration ash. (author)

Method for processing powdery radioactive wastes

International Nuclear Information System (INIS)

Yasumura, Keijiro; Matsuura, Hiroyuki; Tomita, Toshihide; Nakayama, Yasuyuki.

1978-01-01

Purpose: To solidify radioactive wastes with ease and safety at a high reaction speed but with no boiling by impregnating the radioactive wastes with chlorostyrene. Method: Beads-like dried ion exchange resin, powdery ion exchange resin, filter sludges, concentrated dried waste liquor or the like are mixed or impregnated with a chlorostyrene monomer dissolving therein a polymerization initiator such as methyl ethyl ketone peroxide and benzoyl peroxide. Mixed or impregnated products are polymerized to solid after a predetermined of time through curing reaction to produce solidified radioactive wastes. Since inflammable materials are used, this process has a high safety. About 70% wastes can be incorporated. The solidified products have a strength as high as 300 - 400 kg/cm 3 and are suitable to ocean disposal. The products have a greater radioactive resistance than other plastic solidification products. (Seki, T.)

Physicochemical characterization of solidification agents used and products formed with radioactive wastes at LWR nuclear power plants

International Nuclear Information System (INIS)

Kibbey, A.H.; Godbee, H.W.

1978-01-01

Solidification of evaporator concentrates, filter sludges, and spent ion exchange resins used in LWR streams is discussed. The introduction of solidification agents to immobilize these sludges and resins can increase the volume of these wastes by a factor of slightly over 1 to greater than 2, depending on the binder chosen. The agents and methods used or proposed for use in solidification of LWR power plant wastes are generally suitable for treating most of the other-than-high-level wastes generated throughout the entire fuel cycle. Among the solidification agents most commonly used or suggested for use are the inorganic cements and organic plastics, which are listed and compared. A summary of considerations important in choosing a solidification agent is presented tabularly

Radioactive wastes: sources, treatment, and disposal

International Nuclear Information System (INIS)

Wymer, R.G.; Blomeke, J.O.

1975-01-01

Sources, treatment, and disposal of radioactive wastes are analyzed in an attempt to place a consideration of the problem of permanent disposal at the level of established or easily attainable technology. In addition to citing the natural radioactivity present in the biosphere, the radioactive waste generated at each phase of the fuel cycle (mills, fabrication plants, reactors, reprocessing plants) is evaluated. The three treatment processes discussed are preliminary storage to permit decay of the short-lived radioisotopes, solidification of aqueous wastes, and partitioning the long-lived α emitters for separate and long-term storage. Dispersion of radioactive gases to the atmosphere is already being done, and storage in geologically stable structures such as salt mines is under active study. The transmutation of high-level wastes appears feasible in principle, but exceedingly difficult to develop

Mobile concrete solidification systems for power reactor waste

International Nuclear Information System (INIS)

Tchemitcheff, E.; Bordas, Y.

1990-01-01

In late 1988 SGN received an order from Electricite de France (EDF) for the construction of a mobile concrete solidification system to process secondary system resins generated by the P'4 and N4 series PWR power plants in France. This order was placed in view of SGN's experience with low- and medium-level radioactive waste treatment and conditioning over a period of almost 20 years. In addition to the construction of fixed waste processing facilities using more conventional technologies, SGN has been involved in application of the mobile system concept to the bituminization process in the United States, which led to the construction and commissioning of two transportable systems in collaboration with its American licensee US Ecology. It has also conducted large-scale R ampersand D on LLW/MLW concrete solidification, particularly for ion exchange resins. 5 figs

Solidification technique of radioactive elements. Research using zirconium phosphates

International Nuclear Information System (INIS)

Nakayama, Susumu; Ito, Katsuhiko

2005-01-01

Proton type zirconium phosphates HZr 2 (PO 4 ) 3 , NASICON type three-dimensional net work structure, is used for solidification of Cs in the high level radioactive waste. Two kinds of solidification methods such as the dry method and autoclave method are explained. Cs ion entered into 0.6nm space of HZr 2 (PO 4 ) 3 , and formed ionic bonding, which made the difficult situation to remove. When mixture of HZr 2 (PO 4 ) 3 and 23 kinds of M(NO 3 )n (M= Li, Na, K, Pb, Sr, Bi, Y, Mg, Ca, Sc, Mn, Fe, Co, Ni, Cu, Zn, Ag, Cd, Ba, La, Ce, Tl, and Pb; n=1,2 or 3) was treated at 400-700degC by dry method, solidification of the subject metals was succeeded. Amount of solidification of Cs by autoclave at 250degC is almost same as the dry method and its leachability resistance increased 40 times than that of dry method after heat treatment in atmosphere at 700degC. (S.Y.)

Radioactive waste management

International Nuclear Information System (INIS)

Alfredson, P.G.; Levins, D.M.

1975-08-01

Present and future methods of managing radioactive wastes in the nuclear industry are reviewed. In the stages from uranium mining to fuel fabrication, the main purpose of waste management is to limit and control dispersal into the environment of uranium and its decay products, particularly radium and radon. Nuclear reactors produce large amounts of radioactivity but release rates from commercial power reactors have been low and well within legal limits. The principal waste from reprocessing is a high activity liquid containing essentially all the fission products along with the transuranium elements. Most high activity wastes are currently stored as liquids in tanks but there is agreement that future wastes must be converted into solids. Processes to solidify wastes have been demonstrated in pilot plant facilities in the United States and Europe. After solidification, wastes may be stored for some time in man-made structures at or near the Earth's surface. The best method for ultimate disposal appears to be placing solid wastes in a suitable geological formation on land. (author)

Proceedings of the 1st workshop on radioactive waste treatment technologies, October 28, 1997 Taejon, Korea

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

This proceedings describes the volume reduction of radioactive waste, the radioactive waste treatment technology, the decontamination and decommissioning, and the incineration and solidification of radioactive waste. Twenty two papers are submitted.

Proceedings of the 1st workshop on radioactive waste treatment technologies, October 28, 1997 Taejon, Korea

International Nuclear Information System (INIS)

1997-01-01

This proceedings describes the volume reduction of radioactive waste, the radioactive waste treatment technology, the decontamination and decommissioning, and the incineration and solidification of radioactive waste. Twenty two papers are submitted

Treatment methods for radioactive mixed wastes in commercial low-level wastes: technical considerations

International Nuclear Information System (INIS)

MacKenzie, D.R.; Kempf, C.R.

1986-01-01

Treatment options for the management of three generic categories of radioactive mixed waste in commercial low-level wastes (LLW) have been identified and evaluated. These wastes were characterized as part of a BNL study in which LLW generators were surveyed for information on potential chemical hazards in their wastes. The general treatment options available for mixed wastes are destruction, immobilization, and reclamation. Solidification, absorption, incineration, acid digestion, wet-air oxidation, distillation, liquid-liquid wastes. Containment, segregation, decontamination, and solidification or containment of residues, have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, wet-air oxidation, acid digestion, and containment have been considered. For each of these wastes, the management option evaluation has included an assessment of testing appropriate to determine the effect of the option on both the radiological and potential chemical hazards present

Treatment methods for radioactive mixed wastes in commercial low-level wastes: technical considerations

Energy Technology Data Exchange (ETDEWEB)

MacKenzie, D.R.; Kempf, C.R.

1986-01-01

Treatment options for the management of three generic categories of radioactive mixed waste in commercial low-level wastes (LLW) have been identified and evaluated. These wastes were characterized as part of a BNL study in which LLW generators were surveyed for information on potential chemical hazards in their wastes. The general treatment options available for mixed wastes are destruction, immobilization, and reclamation. Solidification, absorption, incineration, acid digestion, wet-air oxidation, distillation, liquid-liquid wastes. Containment, segregation, decontamination, and solidification or containment of residues, have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, wet-air oxidation, acid digestion, and containment have been considered. For each of these wastes, the management option evaluation has included an assessment of testing appropriate to determine the effect of the option on both the radiological and potential chemical hazards present.

Method for solidifying liquid radioactive wastes

International Nuclear Information System (INIS)

Berreth, J.R.

1976-01-01

The quantity of nitrous oxides produced during the solidification of liquid radioactive wastes containing nitrates and nitrites can be substantially reduced by the addition to the wastes of a stoichiometric amount of urea which, upon heating, destroys the nitrates and nitrites, liberating nontoxic N 2 , CO 2 and NH 3 . 5 claims, no drawings

Sandia solidification process: a broad range aqueous waste solidification method

International Nuclear Information System (INIS)

Lynch, R.W.; Dosch, R.G.; Kenna, B.T.; Johnstone, J.K.; Nowak, E.J.

1976-01-01

New ion-exchange materials of the hydrous oxide type were developed for solidifying aqueous radioactive wastes. These materials have the general formula M[M'/sub x/O/sub y/H/sub z/]/sub n/, where M is an exchangeable cation of charge +n and M' may be Ti; Nb; Zr, or Ta. Affinities for polyvalent cations were found to be very high and ion-exchange capacities large (e.g., 4.0--4.5 meq/g for NaTi 2 O 5 H depending on moisture content). The effectiveness of the exchangers for solidifying high-level waste resulting from reprocessing light-water reactor fuel was demonstrated in small-scale tests. Used in conjunction with anion exchange resin, these materials reduced test solution radioactivity from approximately 0.2 Ci/ml to as low as approximately 2 nCi/ml. The residual radioactivity was almost exclusively due to 106 Ru and total α-activity was only a few pCi/ml. Alternative methods of consolidating the solidified waste were evaluated using nonradioactive simulants. Best results were obtained by pressure-sintering which yielded essentially fully dense ceramics, e.g., titanate/titania ceramics with bulk density as high as 4.7 g/cm 3 , waste oxide content as high as 1.2 g/cm 3 , and leach resistance comparable to good borosilicate glass. Based on the above results, a baseline process for solidifying high-level waste was defined and approximate economic analyses indicated costs were not prohibitive. Additional tests have demonstrated that, if desired, operating conditions could be modified to allow recovery of radiocesium (and perhaps other isotopes) during solidification of the remaining constituents of high-level waste. Preliminary tests have also shown that these materials offer promise for treating tank-stored neutralized wastes

Radioactive waste processing container

International Nuclear Information System (INIS)

Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

1992-01-01

A radioactive waste processing container used for processing radioactive wastes into solidification products suitable to disposal such as underground burying or ocean discarding is constituted by using cements. As the cements, calcium sulfoaluminate clinker mainly comprising calcium sulfoaluminate compound; 3CaO 3Al 2 O 3 CaSO 4 , Portland cement and aqueous blast furnace slug is used for instance. Calciumhydroxide formed from the Portland cement is consumed for hydration of the calcium sulfoaluminate clinker. According, calcium hydroxide is substantially eliminated in the cement constituent layer of the container. With such a constitution, damages such as crackings and peelings are less caused, to improve durability and safety. (I.N.)

Method of solidifying and disposing radioactive waste plastic

International Nuclear Information System (INIS)

Matsuura, Hiroyuki; Yasumura, Keijiro

1981-01-01

Purpose: To solidify radioactive waste as it is with plastic by forming a W/O (Water-in-Oil) emulsion with the radioactive waste and a plastic solidifier, and treating it with a polymerization starting agent, an accelerator, and the like. Method: A predetermined amount of alkaline substance such as sodium hydroxide, triethanol, or the like is added quantitatively to radioactive waste and it is mixed by an agitator. A predetermined amount of solidifier such as unsaturated polyester or the like is added to the mixture and it is further mixed by the agitator to form a stable W/O emulsion. Subsequently, predetermined amounts of polymerization starting agent such as methyl ethyl ketone peroxide and polymerization accelerator such as cobalt naphthenate or the like are added thereto, the mixture is mixed, and is then allowed to stand for at room temperature for the plastic solidification thereof. No reaction occurs after the solidification. (Sekiya, K.)

Cementation of biodegraded radioactive oils and organic waste

International Nuclear Information System (INIS)

Gorbunova, O.; Safonov, A.; Tregubova, V.; German, K.

2015-01-01

The possibility of the microbiological pre-treatment of the oil-containing organic liquid radioactive waste (LRW) before solidification in the cement matrix has been studied. It is experimentally proved that the oil containing cement compounds during long-term storage are subject to microbiological degradation due to the reaction of biogenic organic acids with the minerals of the cement matrix. We recommend to biodegrade the LRW components before their solidification, which reduces the volume of LRW and prevent the destruction of the inorganic cement matrix during the long term storage. The biodegradation of the oil containing LRW is possible by using the radioresistant microflora which oxidize the organic components of the oil to carbon dioxide and water. Simultaneously there is the bio-sorption of the radionuclides by bacteria and emulsification of oil in cement slurry due to biogenic surface-active substances of glycolipid nature. It was experimentally established that after 7 days of biodegradation of oil-containing liquid radioactive waste the volume of LRW is reduced by the factor from 2 to 10 due to the biodegradation of the organic phase to the non-radioactive gases (CH 4 , H 2 O, CO 2 , N 2 ), which are excluded from the volume of the liquid radioactive waste. At the same time, the microorganisms are able to extract from the LRW up to 80-90% of alpha-radionuclides, up to 50% of 90 Sr, up to 20% of 137 Cs due to sorption processes at the cellular structures. The radioactive biomass is subject to dehydration and solidification in the matrix. The report presents the following experimental data: type of bacterial flora, the parameters of biodegradation, the cementing parameters, the properties of the final cement compound with oil-containing liquid radioactive waste

Disposal of high-level radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Costello, J M [Australian Atomic Energy Commission Research Establishment, Lucas Heights

1982-03-01

The aims and options for the management and disposal of highly radioactive wastes contained in spent fuel from the generation of nuclear power are outlined. The status of developments in reprocessing, waste solidification and geologic burial in major countries is reviewed. Some generic assessments of the potential radiological impacts from geologic repositories are discussed, and a perspective is suggested on risks from radiation.

Treatment methods for radioactive mixed wastes in commercial low-level wastes - technical considerations

International Nuclear Information System (INIS)

MacKenzie, D.R.; Kempf, C.R.

1986-01-01

Treatment options for the management of three generic categories of radioactive mixed waste in commercial low-level wastes (LLW) have been identified and evaluated. These wastes were characterized as part of a BNL study in which LLW generators were surveyed for information on potential chemical hazards in their wastes. The general treatment options available for mixed wastes are destruction, immobilization, and reclamation. Solidification, absorption, incineration, acid digestion, wet-air oxidation, distillation, liquid-liquid solvent extraction, and specific chemical destruction techniques have been considered for organic liquid wastes. Containment, segregation, decontamination, and solidification or containment of residues, have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, wet-air oxidation, acid digestion, and containment have been considered. Fore each of these wastes, the management option evaluation has included an assessment of testing appropriate to determine the effect of the option on both the radiological and potential chemical hazards present

Solidification and vitrification life-cycle economics study

International Nuclear Information System (INIS)

Gimpel, R.F.

1992-01-01

Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450 000 cm 3 of low-level radioactive and mixed wastes at the Fernald Environmental Management Project (FEMP) located near Cincinnati, Ohio. This paper summarizes a detailed study done to: (1) compare the economics of the solidification and vitrification processes, (2) determine if the stigma assigned to vitrification is warranted and, (3) determine if investing millions of dollars into vitrification development, along with solidification development, at Fernald is warranted

OVERVIEW OF THE HISTORY, PRESENT STATUS, AND FUTURE DIRECTION OF SOLIDIFICATION/STABILIZATION TECHNOLOGIES FOR HAZARDOUS WASTE TREATMENT

Science.gov (United States)

Solidification/stabilization (S/S) technology processes are currently being utilized in the United States to treat inorganic and organic hazardous waste and radioactive waste. These wastes are generated from operating industry or have resulted from the uncontrolled management of ...

Spanish program on disposal of radioactive wastes

International Nuclear Information System (INIS)

Lopez Perez, B.; Ramos Salvador, L.; Martines Martinez, A.

1977-01-01

The Spanish Energetic Program assumes an installed nuclear electrical power of 23.000 MWe by the year 1985. Therefore, Spain is making an effort in the managment of radioactive wastes, that can be synthesized in the following points: 1.- Make-up and review of the regulation on the management of radioactive wastes. 2.- Development of the processes and equipment for the treatment of solid, liquid and gaseous wastes from the CNEN ''Juan Vigon'', as well as those from the Nuclear Center of Soria. Solidification studies of RAA wastes arisen from the reprocessing. 3.- Evaluation of radioactive waste treatment systems of the new installed nuclear power plants. Assistance to the nuclear and radioactive facilities operators. 4.- Increase the storage capacity of the pilot repository for solid radioactive wastes of categories 1 and 2 IAEA, located in Sierra Albarrana. Studies of adequate geological formation for storage of solid wastes of IAEA categories 3 and 4. 5.- Studies about long term surface storage systems for solidified RAA wastes arisen from the reprocessing [es

Method for solidification and disposal of radioactive pellet waste

International Nuclear Information System (INIS)

Yasumura, Keijiro; Matsuura, Hiroyuki.

1975-01-01

Object: To form radioactive waste into pellet, which is impregnated with plastic monomer for polymerization, and then packed into a drum can to have gaps between composites filled with cement, mortar, and molten asphalt, thus increasing water resistance and strength. Structure: Radioactive powdery bodies discharged from a thin film scaraping drier are formed into pellets in the desired shape. The thus pelletized radioactive solid waste is impregnated with a fluid plastic monomer such as styrene monomer and methacrylacidmethyl, and a polymerization accelerator is added thereto for polymerization. As a consequence, a composite pellet of powdery solid waste and plastic may be obtained. This is packed into the drum can container, into which cement paste, cement mortar or molten asphalt are put to fill the space between the plastic pellet composites, thus obtaining a solidified body integral with the drum can. (Taniai, N.)

Management of high level radioactive waste

International Nuclear Information System (INIS)

Redon, A.; Mamelle, J.; Chambon, M.

1977-01-01

The world wide needs in reprocessing will reach the value of 10.000 t/y of irradiated fuels, in the mid of the 80's. Several countries will have planned, in their nuclear programme, the construction of reprocessing plants with a 1500 t/y capacity, corresponding to 50.000 MWe installed. At such a level, the solidification of the radioactive waste will become imperative. For this reason, all efforts, in France, have been directed towards the realization of industrial plants able of solidifying the fission products as a glassy material. The advantages of this decision, and the reasons for it are presented. The continuing development work, and the conditions and methods of storing the high-level wastes prior to solidification, and of the interim storage (for thermal decay) and the ultimate disposal after solidification are described [fr

Leaching behavior of solidified plastics radioactive wastes

International Nuclear Information System (INIS)

Yook, Chong Chul; Lee, Byung Hun; Jae, Won Mok; Kim, Kyung Eung

1986-01-01

It is highly needed to develope the solidification process to dispose safely the radioactive wastes increasing with the growth of the nuclear industry. The leaching mechanisms of the solidified plastic wastes were investigated and the leaching rates of the plastic wastes were also measured among the many solidification processes. In addition, the transport equation based on the diffusion or the diffusion-dissolution was compared with the empirical equation derived from the experimental data by graphical method. Consequently, leaching process of the solidified plastic wastes is quite well agreed with the mass transport theory, but it may be difficult to simulate leaching process by diffusion dissolution mechanism. But the theoretical equation could be applicable to the cumulative amount of radionuclides leached form the plastic wastes disposed into the environment. (Author)

Method of solidifying radioactive laundry wastes

International Nuclear Information System (INIS)

Yasumura, Keijiro

1984-01-01

Purpose: To enable to solidify radioactive laundry wastes containing non-ionic liquid detergents less solidifiable by plastic solidification process in liquid laundry wastes for cloths or the likes discharged from a nuclear power plant. Method: Radioactive laundry wastes are solidified by using plastic solidifying agent comprising, as a main ingredient, unsaturated polyester resins and methylmethacrylate monomers. The plastic solidifying agents usable herein include, for example, unsaturated polyester resins prepared by condensating maleic anhydride and phthalic anhydride with propylene glycol and incorporated with methylmethacrylate monomers. The mixing ratio of the methylmethacrylate monomers is preferably 30 % by weight based on the unsaturated polyester resins. (Aizawa, K.)

Treatment of radioactive mixed wastes in commercial low-level wastes

International Nuclear Information System (INIS)

Kempf, C.R.; MacKenzie, D.R.

1985-01-01

Management options for three generic categories of radioactive mixed waste in commercial low-level wastes have been identified and evaluated. These wastes were characterized as part of a BNL study in which a large number of generators were surveyed for information on potentially hazardous low-level wastes. The general management targets adopted for mixed wastes are immobilization, destruction, and reclamation. It is possible that these targets may not be practical for some wastes, and for these, goals of stabilization or reduction of hazard are addressed. Solidification, absorption, incineration, acid digestion, segregation, and substitution have been considered for organic liquid wastes. Containment, segregation, and decontamination and re-use have been considered for lead metal wastes which have themselves been contaminated and are not used for purposes of waste disposal shielding, packaging, or containment. For chromium-containing wastes, solidification, incineration, containment, substitution, chemical reduction, and biological removal have been considered. For each of these wastes, the management option evaluation has necessarily included assessment/estimation of the effect of the treatment on both the radiological and potential chemical hazards present. 10 refs

Interim solidification of SRP waste with silica, bentonite, or phosphoric acid

International Nuclear Information System (INIS)

Thompson, G.H.

1976-03-01

One option for interim waste management at the Savannah River Plant is in-tank solidification of the liquid waste solutions. This would reduce the mobility of these highly radioactive solutions until techniques for their long-term immobilization and storage are developed and implemented. Interim treatments must permit eventual retrieval of waste and subsequent incorporation into a high-integrity form. This study demonstrated the solidification of simulated alkaline waste solutions by reaction with silica, bentonite, and phosphoric acid. Alkaline waste can be solidified by reaction with silica gel, silica flour, or sodium silicate solution. Solidified products containing waste salt can be retrieved by slurrying with water. Alkaline supernate (solution in equilibrium with alkaline sludge in SRP waste tanks) can be solidified by reaction with bentonite to form cancrinite powder. The solidified waste can be retrieved by slurrying with water. Alkaline supernate can be solidified by partial evaporation and reaction with phosphoric acid. Water is incorporated into hydrated complexes of trisodium phosphate. The product is soluble, but actual plant waste would not solidify completely because of decay heat. Reaction of simulated alkaline waste solutions with silica gel, silica flour, or bentonite increases the volume by a factor of approximately 6 over that of evaporated waste; reaction with phosphoric acid results in a volume 1.5 times that of evaporated waste. At present, the best method for in-tank solidification is by evaporation, a method that contributes no additional solids to the waste and does not compromise any waste management options

MethodS of radioactive waste processing and disposal in the United Kingdom

International Nuclear Information System (INIS)

Tolstykh, V.D.

1983-01-01

The results of investigations into radioactive waste processing and disposal in the United Kingdom are discussed. Methods for solidification of metal and graphite radioactive wastes and radioactive slime of the Magnox reactors are described. Specifications of different installations used for radioactive waste disposal are given. Climatic and geological conditions in the United Kingdom are such that any deep storages of wastes will be lower than the underground water level. That is why dissolution and transport by underground waters will inevitably result in radionuclide mobility. In this connection an extended program of investigations into the main three aspects of disposal problem namely radionucleide release in storages, underground water transport and radionuclide migration is realized. The program is divided in two parts. The first part deals with retrival of hydrological and geochemical data on geological formations, development of specialized methods of investigations which are necessary for identification of places for waste final disposal. The second part represents theoretical and laboratory investigations into provesses of radionuclide transport in the system of ''sttorage-geological formation''. It is concluded that vitrification on the base of borosilicate glass is the most advanced method of radioactive waste solidification

Nuclear waste solidification

Science.gov (United States)

Bjorklund, William J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition.

Nuclear waste solidification

International Nuclear Information System (INIS)

Bjorklund, W.J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition

Development of radioactive waste treatment system for nuclear power stations by Toshiba (III)

International Nuclear Information System (INIS)

Irie, H.; Takahara, T.; Matsuda, T.; Matsuura, H.; Yasumura, K.; Nakayama, Y.

1989-01-01

This paper describes a solidification process with thermosetting resin to satisfy both requirements of volume reduction and quality of solidified products. Volumes of solidified products in drums generated from spent resins and concentrated wastes were reduced respectively to 1/4 and less than 1/6 of those in the conventional cement solidification process. In plants using a simple demineralizing system for condensate polishing, a large amount of waste water with regenerant chemicals is generated from the condensate demineralizer. In general, radioactivity concentration of wastes from this type of nuclear power plant is comparatively high, so the dose rate at the surface of drums containing solidified wastes exceeds 200mR/h. A pelletizing system for radioactive wastes was developed to reduce their volumes and allow their interim storage until the radioactivity decays down to a level at which they can be handled easily

Processing of concentrated radioactive wastes into cement and bitumens following calcination

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Ditl, P.; Prikryl, P.

1985-01-01

A brief characteristic is presented of the most frequently used processes of solidification of liquid radioactive wastes, viz., bituminization, cementation and their combination with calcination. The effect of individual parameters is assessed on the choice of the type of solidification process as is their importance in the actual process, in temporary storage, during transportation and under conditions of long-term storage. It has been found that a combination of the procedures could lead to a modular system of methods and equipment. This would allow to approach optimal solidification of wastes in the present period and to establish a research reserve for the development of more modern, economically advantageous and safer procedures. A rough estimate is made of the costs of the solidification of 1 m 3 of radioactive concentrate from the V-1 power plant at a production of 380 m 3 /year, this for the cementation-calcination and bituminization-calcination procedures. The said rough economic analysis only serves to identify the major operating components which have the greatest effect on the economic evaluation of the solidification procedures. (Z.M.)

Solidification of radioactive aqueous solution

Energy Technology Data Exchange (ETDEWEB)

Aikawa, Hideaki; Kato, Kiyoshi; Wadachi, Yoshiki

1970-09-07

A process for solidifying a radioactive waste solution is provided, using as a solidifying agent a mixture of calcined gypsum and burnt vermiculite. The quantity ratio of the mixture is preferred to be 1:1 by volume. The quantity of impregnation is 1/2 of the volume of the total quantity of the solidifying agent. In embodiments, 10 liters of plutonium waste solution was mixed with a mixture of 1:1 calcined gypsum and burnt vermiculite contained in a 20-liter cylindrical steel container lined with asphalt. The plutonium waste solution from the laboratory was neutralized with a caustic soda aqueous solution to prevent explosion due to the nitration of organic compounds. The neutralization is not always necessary. A market available dental gypsum was calcined at 400 to 500/sup 0/C and a vermiculite from Illinois was burnt at 1,100/sup 0/C to prepare the agents. The time required for the impregnation with 10 liters of plutonium solution was four minutes. After impregnation, the temperature rose to 40/sup 0/C within 30 minutes to one hour. Next, it was cooled to room temperature by standing for 3-4 hours. Solidification time was about 1 hour. The Japan Atomic Energy Research Insitute had treated and disposed about 1,000 tons of plutonium waste by this process as of August 19, 1970.

Immobilization of radioactive waste in glass matrices

International Nuclear Information System (INIS)

Wicks, G.G.

1978-01-01

A promising process for long-term management of high-level radioactive waste is to immobilize the waste in a borosilicate glass matrix. Among the most important criteria characterizing the integrity of the large-scale glass-waste forms are that they possess good chemical stability (including low leachability), thermal stability, mechanical integrity, and high radiation stability. Fulfillment of these criteria ensures the maximum margin of safety of glass-waste products, following solidification, handling, transportation, and long-term storage

Sponsored research on radioactive waste management. Progress report January 1981 - March 1982

Energy Technology Data Exchange (ETDEWEB)

1983-01-01

The report is in chapters entitled: introduction (background, responsibilities, options, structure of the programme); strategy development; disposal of accumulations; disposal of radioactive waste arisings; quality assurance for waste conditioning quality assurance related to radioactive waste disposal (effectiveness of different rock types as natural barriers to the movement of radioactivity, and non-site specific factors in the design of repositories; radiological assessment; environmental studies; research and development to meet requirements specific to UKAEA wastes; long term research (processes for the solidification of highly active liquid wastes); plutonium contamination waste minimisation.

Removal of nitrogen oxides, 106RuO4 vapors and radioactive aerosols from the gas originating in radioactive wastes solidification

International Nuclear Information System (INIS)

Kepak, F.; Pecak, V.; Uher, E.; Kanka, J.; Koutova, S.; Matous, V.

1985-01-01

Procedures and equipment for the disposal of nitrogen oxides, RuO 4 vapors and radioactive aerosols of 90 Sr, 137 Cs, 60 Co and 125 Sb contained in the gas generated in the solidification of high- and intermediate-level radioactive wastes were tested on models. Nitrogen oxides were disposed of by absorption and chemical decomposition in various solutions of which the best results gave solutions of ammonium salts. Absorption in solutions, physical and chemical sorption on inorganic sorbents were tested for the disposal of RuO 4 . Aerosols were disposed of by absorption in absorption media with subsequent filtration. Of fibrous filter materials, Czechoslovak AEROS-2 and RA-2 filter papers were proven in the tests. Attention was also devoted to granular filter materials of which silica gel was chosen. On the basis of laboratory tests a multi-step treatment system was designed which consists of a condenser, a nitrogen oxide absorber, a liquid aerosol separator, absorption columns and aerosol filters. The whole system has been manufactured on pilot plant scale and the different parts are being produced. (Z.M.)

Solidification of radioactive waste effluents

International Nuclear Information System (INIS)

Mergan, L.M.; Cordier, J.-P.

1981-01-01

A process and apparatus for solidifying radioactive waste liquid containing dissolved and/or suspended solids is disclosed. The process includes chemically treating for pH adjustment and precipitation of solids, concentrating solids with a thin-film evaporator to provide liquid concentrate containing about 50% solids, and drying the concentrate with a heated mixing apparatus. The heated mixing apparatus includes a heated wall and working means for shearing dried concentrate from internal surfaces and subdividing dry concentrate into dry, powdery particles. The working means includes a rotor and helical means for positively advancing the concentrate and resulting dry particles from inlet to outlet of the mixing apparatus. The dry particles may also be encapsulated in a matrix material. Entrained particles in the vapor stream from the evaporator and mixer are removed in an integral particle separator and the vapor is subsequently condensed and may be recycled upstream of the thin-film evaporator. A section of the mixer may be used for mixing dry particles with the matrix material in a continuous drying and mixing sequence. A section of the mixer also may be used for mixing the treating chemical with the waste liquid

Fixation process for radioactive waste

International Nuclear Information System (INIS)

Theysohn, F.

1977-01-01

An improvement on the method of solidification of radioactive liquid waste in bitumen with the aid of extruders is described. So far, it has been difficult to remove large amounts of water. The waste sludge, as proposed here, is pre-dried in the extruder and then mixed with the bitumen. The extruder is inclined upward in the transport direction, and its barrel extruders have through holes parallel to the direction of transport in the raised sides of the passages, so that water runs back. Also the waste steam nozzles are arranged before the bitumen inlet. (UWI) [de

Solidification of radioactive wastes with thermosetting resin

International Nuclear Information System (INIS)

Hayashi, M.; Kobayashi, K.; Okamoto, O.; Kagawa, T.; Wakamatsu, K.; Irie, H.; Matsuura, H.; Yasumura, K.; Nakayama, Y.

1982-01-01

Dried simulated radioactive wastes were solidified with thermosetting resin and their properties were investigated with laboratory scale and real scale products through extensive testings, such as mechanical resistance, resistance to leaching and swelling in water, radiation resistance, fire resistance and resistance to temperature cycling. The typical results were as follows: over 600 kg/cm 2 of compressive strength, diffusion constant of approx. 10 - 5 cm 2 /day for 137 Cs leaching from solidified waste products, no significant change was found for up to 5 x 10 8 RAD irradiation, and damages were limited to the surface of the products after the thermal test and dropping impact test. 7 figures, 4 tables

Influence of non-technical policies on choices of waste solidification technologies

International Nuclear Information System (INIS)

Trubatch, S.L.

1987-01-01

This paper describes and discusses non-technical policy considerations which may improperly influence decisions on the solidification of low-level radioactive wastes (''LLW''). These policy considerations are contained principally in several State and Federal statutes which regulate various aspects of LLW disposal. One policy consideration in particular, the unqualified bias in favor of volume reduction, is shown to present a substantial potential for leading to technically suboptimal decisions on the appropriate processes for solidifying LLW. To avoid the unintended skewing of technical decisions by non-technical policy considerations, certain current policies may need to be revised to ensure that the choices of waste treatment, including decisions on solidification, are based primarily on reasonable assurance of adequate protection of public health and safety. This goal may be realized in part by basing any disposal fee structure on more than just LLW volume to include consideration of the waste's activity and its difficulty of confinement

Low-level radwaste solidification

International Nuclear Information System (INIS)

Naughton, M.D.; Miller, C.C.; Nelson, R.A.; Tucker, R.F.

1983-01-01

This paper reports on a study of ''Advanced Low-Level Radioactive Waste Treatment Systems'' conducted under an EPRI contract. The object of the study is to identify advanced lowlevel radwaste treatment systems that are commercially available or are expected to be in the near future. The current state-ofthe-art in radwaste solidification technology is presented. Related processing technologies, such as the compaction of dry active waste (DAW), containers available for radwaste disposal, and the regulatory aspects of radwaste transportation and solidification, are described. The chemical and physical properties of the currently acceptable solidification agents, as identified in the Barnwell radwaste burial site license, are examined. The solidification agents investigated are hydraulic cements, thermoplastic polymers, and thermosetting polymers. It is concluded that solidification processes are complex and depend not only on the chemical and physical properties of the binder material and the waste, but also on how these materials are mixed

Discarding processing method for radioactive waste

International Nuclear Information System (INIS)

Komura, Shiro; Kato, Hiroaki; Hatakeyama, Takao; Oura, Masato.

1992-01-01

At first, in a discrimination step, extremely low level radioactive wastes are discriminated to metals and concretes and further, the metal wastes are discriminated to those having hollow portions and those not having hollow portions, and the concrete wastes are discriminated to those having block-like shape and those having other shapes respectively. Next, in a processing step, the metal wastes having hollow portions are applied with cutting, devoluming or packing treatment and block-like concrete wastes are applied with surface solidification treatment, and concrete wastes having other shapes are applied with crushing treatment respectively. Then, the extremely low level radioactive wastes contained in a container used exclusively for transportation are taken out, in a movable burying facility with diffusion inhibiter kept at a negative pressure as required, in a field for burying operation, and buried in a state that they are isolated from the outside. Accordingly, they can be buried safely and efficiently. (T.M.)

Liquid radioactive waste processing system for pressurized water reactor plants

International Nuclear Information System (INIS)

Anon.

1976-01-01

This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

Boiling water reactor liquid radioactive waste processing system

International Nuclear Information System (INIS)

Anon.

1977-01-01

The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

Solidification of liquid radioactive concentrates by fixation with cement

International Nuclear Information System (INIS)

Pekar, A.; Breza, M.; Timulak, J.; Krajc, T.

1985-01-01

In testing the technology of liquid radioactive wastes cementation, the effect was mainly studied of the content of boric acid and its salts on cement solidification, the effect of additives on radionuclide leachability and the effect of the salt content on the cementation product. On the basis of experimental work carried out on laboratory scale with model samples and samples of radioactive concentrate from the V-1 nuclear power plant, the following suitable composition of the cementation mixture was determined: 40% Portland cement, 40% zeolite containing material and 20% power plant ash. The most suitable ratio of liquid radioactive wastes and the cementation mixture is 0.5. As long as in such case the salt content of the concentrate ranges between 20 and 25%, the cementation product will have a maximum salt content of 10% and a leachability of the order of 10 -3 to 10 -4 g/cm 2 per day with a mechanical strength allowing safe handling. It was also found that the quality processing of the cement paste with degassing, e.g., by vibration, is more effective for the production of a pore-free cementation product than the application of various additives which are supposed to eliminate pore formation. (Z.M.)

Analysis by Fourier Transform Infrared (FTIR) of the gamma radiation effect on epoxy resin, used as solidification agent of radioactive wastes

International Nuclear Information System (INIS)

Liu, C.H.; Riella, H.G.; Guedes, S.M.L.

1995-01-01

The effects of gamma radiation on Epoxy resin, used as solidification agent of radioactive wastes, were studied by Fourier Transform Infrared (FTIR). The spectra showed no significant modifications on Epoxy resin functional groups (irradiated with dose from 0 to 1 MGy). Up to 1 MGy Epoxy resin did not oxidize, confirming the Epoxy good radiation strength. The presence of aromatic chain and amine group, mainly tertiary amine, give good radiolytic stability to the Epoxy, increasing the interest to use this material in nuclear facilities. (author). 3 refs, 2 figs

Solidification and vitrification life-cycle economics study

International Nuclear Information System (INIS)

Gimpel, R.F.

1992-01-01

Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method, whereas vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ex situ solidification and vitrification are the competing methods for treating in excess of 450,000 m 3 of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper s a detailed study done to: compare the economics of the solidification and vitrification processes; determine if the stigma assigned to vitrification is warranted; determine if investing millions of dollars into vitrification development, along with solidification development, at Fernald is warranted. Common parameters were determined and detailed life-cycle cost estimates were made. Incorporating the unit costs into a computer spreadsheet allowed 'what if' scenarios to be performed. Some scenarios investigated included variation of: remediation times, amount of wastes treated, treatment efficiencies, electrical and material costs and escalation

Method of electrolytic processing for radioactive liquid waste

International Nuclear Information System (INIS)

Otsuka, Katsuyuki; Takahashi, Yoshiharu; Tamai, Hideaki.

1989-01-01

Radioactive liquid wastes containing sodium compounds are electrolized using mercury as a cathode. As a result, they are separated into sodium-containing metal amalgam and residues. Metals containing sodium are separated from amalgam, purified and re-utilized, while mercury is recycled to the electrolysis vessel. The foregoing method can provide advantageous effect such as: (1) volume of the wastes to be processed can be reduced, (2) since processing can be carried out at a relatively low temperature, low boiling elements can be handled with no evaporization, (3) useful elements can be recovered and (4) other method than glass solidification can easily be employed remarkable volume-reduction of solidification products can be expected. (K.M.)

Proceedings of the workshop on radioactive, hazardous, and/or mixed waste sludge management

International Nuclear Information System (INIS)

Lomenick, T.F.

1992-01-01

A workshop sponsored by the US Department of Energy (DOE) Field Office, Oak Ridge, was held on December 4--6, 1990, in Knoxville, Tennessee. The primary objective of the workshop was the exchange of information, experiences, solutions, and future plans of DOE and its prime contractors who are engaged in work on the packaging, grouting, storage, and transport of waste sludges. In addition, the group met with industrial participants in an open forum to discuss problems and needs in the management of these wastes and to learn of possible industrial experiences, approaches, and solutions, including demonstrations of potential tools and techniques. Topics discussed include the following: mixed waste sludge issue at the K-25 site; processing saltstone from waste streams at the Savannah River Plant; the Hanford Grout Treatment Facility; treatment of pond sludge at the Rocky Flats Plant; cement solidification of low-level radioactive sludge at the West Valley Demonstration Project; studies on the solidification of low-level radioactive wastes in cement at INEL; cement solidification systems at Los Alamos National Laboratory; emergency avoidance solidification campaign at ORNL; diffusion plant sludge storage problems at the Portsmouth Gaseous Diffusion Plant; the proposed fixation of sludge in cement at the feed materials production center; regulatory aspects of sludge management; and delisting efforts for K-1407-C pond sludges. Individual projects are processed separately for the data bases

Method and device for solidifying radioactive waste

International Nuclear Information System (INIS)

Hayashi, Tadamasa.

1981-01-01

Purpose: To solidify radioactive waste without producing radioactive dusts by always heating and evaporating the water from liquid radioactive waste in a mixture of liquid plastic and exhausting the molten mixture of the waste residue and the plastic material. Constitution: Liquid plastic material in a tank cooled to prevent polymerization or changes of its properties is continuously supplied to the top of a heating and mixing evaporator by a constant supply pump. After the heat transfer surface of the evaporator is covered with the plastic material, radioactive waste in the tank is supplied to the evaporator via the constant supply pump. The waste is abruptly mixed with the plastic material by an agitating rotor, heated by a heater, and the evaporated water is fed to a condenser. An anhydrous molten mixture is continuously exhausted from the bottom of the evaporator into a mixture cooler, a polymerizing agent and catalyst are introduced thereinto from a polymerizing agent tank and a catalyst tank, inhibitor is introduced thereinto from a polymerization inhibitor tank as required, and is filled with the mixture a solidifying container while it is cooled for its polymerization and solidification. (Yoshino, Y.)

Propertis of solidified radioactive wastes from commercial LWRs

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1978-01-01

A study has been performed to characterize the properties of solidified radioactive wastes generated in the liquid radwaste treatment systems at LWRs. The properties which have been studied are those which are pertinent in defining the relative potential for the release of radionuclides to the environment as well as others relating to the evaluation of various solidification agents on an economic and feasibility basis. The use of standard testing procedures in measuring these properties allows an intercomparison of respective properties between various types of solidified waste forms. The leachability, mechanical properties, thermal stability, radiation stability, and thermal properties of hydraulic cement, ureaformaldehyde, bitumen, and addition type polymer waste forms have been measured. In addition, the chemical sensitivity, volumetric efficiency and radiation shielding characteristics of these waste forms have been studied. Emphasis in this paper is placed on the results of studies concerning chemical compatibility of solidification agents with specific waste streams, volumetric efficiency, free standing water, and leachability

Radioactive waste problems in the Kozloduy NPP

Energy Technology Data Exchange (ETDEWEB)

Videnov, N; Stanchev, V [Kombinat Atomna Energetika, Kozloduj (Bulgaria)

1996-12-31

An average volume of 1400 m{sup 3} a year of solid radioactive waste (RAW) is generated in the Kozloduy NPP. The adopted waste processing sequence is collection, sorting and compaction with a 1000 tons force providing decrease in volume by factor of 15. A temporary storage facility at the Kozloduy NPP is licensed by ISUAE and CPPUAE. The treatment of liquid wastes is performed by Westinghouse formula and a technology using an automated solidification system. Contaminated oils are burned using an oil incinerator. A special 2-year programme for RAW management is being developed.

Radioactive waste problems in the Kozloduy NPP

International Nuclear Information System (INIS)

Videnov, N.; Stanchev, V.

1995-01-01

An average volume of 1400 m 3 a year of solid radioactive waste (RAW) is generated in the Kozloduy NPP. The adopted waste processing sequence is collection, sorting and compaction with a 1000 tons force providing decrease in volume by factor of 15. A temporary storage facility at the Kozloduy NPP is licensed by ISUAE and CPPUAE. The treatment of liquid wastes is performed by Westinghouse formula and a technology using an automated solidification system. Contaminated oils are burned using an oil incinerator. A special 2-year programme for RAW management is being developed

Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices; Inertizzazione di rifiuti tossici e nocivi (RTN). Parte seconda: Inertizzazione in matrici cementizie

Energy Technology Data Exchange (ETDEWEB)

Donato, A; Arcuri, L; Dotti, M; Pace, A; Pietrelli, L; Ricci, G [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Casaccia (Italy); Basta, M; Cali, V; Pagliai, V [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Saluggia (Italy)

1989-05-15

This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

Radioactive wastes in nuclear fuel cycle

International Nuclear Information System (INIS)

Sakata, Sadahiro; Nagaike, Tadakatsu; Emura, Satoru; Matsumoto, Akira; Morisawa, Shinsuke.

1978-01-01

Recent topics concerning radioactive water management and disposal are widely reviewed. As the introduction, various sources of radioactivity including uranium mining, fuel fabrication, reactor operation and fuel reprocessing and their amount of wastes accumulated per 1000 MWe year operation of a LWR are presented together with the typical methods of disposal. The second section discusses the problems associated with uranium fuel fabrication and with nuclear power plants. Typical radioactive nuclides and their sources in PWRs and BWRs are discussed. The third section deals with the problems associated with reprocessing facilities and with mixed oxide fuel fabrication. Solidification of high-level wastes and the methods of the disposal of transuranic nuclides are the main topics in this section. The fourth section discusses the methods and the problems of final disposal. Various methods being proposed or studied for the final disposal of low- and high-level wastes and transuranic wastes are reviewed. The fifth section concerns with the risk analysis of waste disposal. Both deterministic and probabilistic methods are treated. As the example, the assessment of the risk due to floods is explained. The associated event tree and fault three are presented together with the estimated probability of the occurrence of each constituent failure. In the final section, the environmental problems of radioactive wastes are widely reviewed. (Aoki, K.)

Plasma separation process: Disposal of PSP radioactive wastes

International Nuclear Information System (INIS)

1989-07-01

Radioactive wastes, in the form of natural uranium contaminated scrap hardware and residual materials from decontamination operations, were generated in the PSP facilities in buildings R1 and 106. Based on evaluation of the characteristics of these wastes and the applicable regulations, the various options for the processing and disposal of PSP radioactive wastes were investigated and recommended procedures were developed. The essential features of waste processing included: (1) the solidification of all liquid wastes prior to shipment; (2) cutting of scrap hardware to fit 55-gallon drums and use of inerting agents (diatomaceous earth) to eliminate pyrophoric hazards; and (3) compaction of soft wastes. All PSP radioactive wastes were shipped to the Hanford Site for disposal. As part of the waste disposal process, a detailed plan was formulated for handling and tracking of PSP radioactive wastes, from the point of generation through shipping. In addition, a waste minimization program was implemented to reduce the waste volume or quantity. Included in this document are discussions of the applicable regulations, the types of PSP wastes, the selection of the preferred waste disposal approach and disposal site, the analysis and classification of PSP wastes, the processing and ultimate disposition of PSP wastes, the handling and tracking of PSP wastes, and the implementation of the PSP waste minimization program. 9 refs., 1 fig., 8 tabs

Nuclear waste disposal: alternatives to solidification in glass proposed

International Nuclear Information System (INIS)

Kerr, R.A.

1979-01-01

More than a quarter-million cubic meters of liquid radioactive wastes are now being held at government installations awaiting final disposal. During the past 20 years, the disposal plan of choice has been to incorporate the 40 to 50 radioactive elements dissolved in liquid wastes into blocks of glass, seal the glass in metal canisters, and insert the canisters into deep, geologically stable salt beds. Over the last few years, some geologists and materials scientists have become concerned that perhaps not enough is known yet about the interaction of waste, container, and salt (or any rock) to have a reasonable assurance that the hazardous wastes will be contained successfully. The biggest advantage of glass at present is the demonstrated practicality of producing large, highly radioactive blocks of it. The frontrunner as a successor to glass is ceramics, which are nonmetallic crystalline materials formed at high temperature, such as chinaware or natural minerals. An apparent advantage of ceramics is that they already have an ordered atomic structure, whose properties can be tailored to a particular waste element and to conditions of a specific disposal site. A ceramic tailored for waste disposal called supercalcine-ceramic has been developed. It was emphasized that the best minerals for waste solidification may be those that have proved most stable under natural conditions over geologic time. Disadvantage to ceramics are radiation damage and transmutation. However, it is now obvious that some ceramics are more stable than glass under certain conditions. Metal-encapsulated ceramic, called cermet, is being developed as a waste form. Cermets are considerably more resistant at 100 0 C than a borosilicate waste glass. Researchers are now testing prospective waste forms under the most extreme conditions that might prevail in a waste disposal site

Low level waste solidification practice in Japan

International Nuclear Information System (INIS)

Sakata, S.; Kuribayashi, H.; Kono, Y.

1981-01-01

Both sea dumping and land isolation are planned to be accomplished for low level waste disposal in Japan. The conceptual design of land isolation facilities has been completed, and site selection will presently get underway. With respect to ocean dumping, safety surveys are being performed along the lines of the London Dumping Convention and the Revised Definitions and Recommendations of the IAEA, and the review of Japanese regulations and applicable criteria is being expedited. This paper discusses the present approach to waste solidification practices in Japan. It reports that the bitumen solidification process and the plastic solidification process are being increasingly used in Japan. Despite higher investment costs, both processes have advantages in operating cost, and are comparable to the cement solidification process in overall costs

Development of sodium disposal technology. Experiment of sodium compound solidification process

International Nuclear Information System (INIS)

Matsumoto, Toshiyuki; Ohura, Masato; Yatoh, Yasuo

2007-07-01

A large amount of sodium containing radioactive waste will come up at the time of final shutdown/decommission of FBR plant. The radioactive waste is managed as solid state material in a closed can in Japan. As for the sodium, there is no established method to convert the radioactive sodium to solid waste. Further, the sodium is highly reactive. Thus, it is recommended to convert the sodium to a stable substance before the solidification process. One of the stabilizing methods is conversion of sodium into sodium hydroxide solution. These stabilization and solidification processes should be safe, economical, and efficient. In order to develop such sodium disposal technology, nonradioactive sodium was used and a basic experiment was performed. Waste-fluid Slag Solidification method was employed as the solidification process of sodium hydroxide solution. Experimental parameters were mixing ratio of the sodium hydroxide and the slag solidification material, temperature and concentration of the sodium hydroxide. The best parameters were obtained to achieve the maximum filling ratio of the sodium hydroxide under a condition of enough high compressive strength of the solidified waste. In a beaker level test, the solidified waste was kept in a long term and it was shown that there was no change of appearance, density, and also the compressive strength was kept at a target value. In a real scale test, homogeneous profiles of the density and the compressive strength were obtained. The compressive strength was higher than the target value. It was shown that the Waste-fluid Slag Solidification method can be applied to the solidification process of the sodium hydroxide solution, which was produced by the stabilization process. (author)

Problems of solidificated radioactive wastes burial into deep geological structures

International Nuclear Information System (INIS)

Kedrovskij, O.L.; Leonov, E.A.; Romadin, N.M.; Shishcits, I.Yu.

1981-01-01

Perspectives are noted of the radioactive wastes burial into deep geopogical structures. For these purposes it has been proposed to investigate severap types of rocks, which do not have intensive gas-generation when beeng heated; salt deposits and clays. Basing on the results of calculations it has been shown that the dimentions of zones of substantial deformations in the case of the high-level radioactive wastes burial to not exceed several hundreds of meters. Conclusion is made that in the case of choosing the proper geotogicat structure for burial and ir the case of inclusion in the structure of the burial site a zone of sanitary alienation, it is possible to isolate wastes safely for all the period of preservation. Preliminary demands have been formulated to geological structures and underground burial sites. As main tasks for optimizatiop of burial sited are considered: determination of necessary types, number and reliability of barriers which ensure isolation of wastes; to make prognoses of the stressed and deformed state of a geological massif on the influence of thermal field; investigation in changes of chemical and physical properties of rocks under heat, radiative and chemical influence; estimation of possible diffusion of radioactivity in a mountin massif; development of a rational mining-thechnological schemes of the burual of wastes of different types. A row of tasks in the farmeworks of this probtem are sotved successfutty. Some resutts are given of the theoretical investigations in determination of zones of distructions of rocks because of heat-load [ru

Radioactive-site-remediation technologies seminar. Speaker slide copies

International Nuclear Information System (INIS)

1992-06-01

The contents of this report include the following: approaches to sampling radioactive heterogeneous waste; soil characterization methodology for determining application of soil washing; vorce (volume reduction/chemical extraction) program; treatment of radioactive compounds in water; polymer solidification of low-level radioactive, hazardous, and mixed waste; in situ vitrification of soils contaminated with radioactive and mixed wastes; decontamination of contaminated buildings; incineration of radioactive waste; in situ stabilization/solidification with cement-based grouts; environmental restoration and waste management; removal of contaminants from soils by electrokinetics; and treatment, compaction, and disposal of residual radioactive waste

Development of a geopolymer solidification method for radioactive wastes by compression molding and heat curing

International Nuclear Information System (INIS)

Shimoda, Chiaki; Matsuyama, Kanae; Okabe, Hirofumi; Kaneko, Masaaki; Miyamoto, Shinya

2017-01-01

Geopolymer solidification is a good method for managing waste because of it is inexpensive as compared with vitrification and has a reduced risk of hydrogen generation. In general, when geopolymers are made, water is added to the geopolymer raw materials, and then the slurry is mixed, poured into a mold, and cured. However, it is difficult to control the reaction because, depending on the types of materials, the viscosity can immediately increase after mixing. Slurries of geopolymers easily attach to the agitating wing of the mixer and easily clog the plumbing during transportation. Moreover, during long-term storage of solidified wastes containing concentrated radionuclides in a sealed container without vents, the hydrogen concentration in the container increases over time. Therefore, a simple method using as little water as possible is needed. In this work, geopolymer solidification by compression molding was studied. As compared with the usual methods, it provides a simple and stable method for preparing waste for long-term storage. From investigations performed before and after solidification by compression molding, it was shown that the crystal structure changed. From this result, it was concluded that the geopolymer reaction proceeded during compression molding. This method (1) reduces the energy needed for drying, (2) has good workability, (3) reduces the overall volume, and (4) reduces hydrogen generation. (author)

Rock solidification method

International Nuclear Information System (INIS)

Nakaya, Iwao; Murakami, Tadashi; Miyake, Takafumi; Funakoshi, Toshio; Inagaki, Yuzo; Hashimoto, Yasuhide.

1985-01-01

Purpose: To convert radioactive wastes into the final state for storage (artificial rocks) in a short period of time. Method: Radioactive burnable wastes such as spent papers, cloths and oils and activated carbons are burnt into ashes in a burning furnace, while radioactive liquid wastes such as liquid wastes of boric acid, exhausted cleaning water and decontaminating liquid wastes are powderized in a drying furnace or calcining furnace. These powders are joined with silicates as such as white clay, silica and glass powder and a liquid alkali such as NaOH or Ca(OH) 2 and transferred to a solidifying vessel. Then, the vessel is set to a hydrothermal reactor, heated and pressurized, then taken out about 20 min after and tightly sealed. In this way, radioactive wastes are converted through the hydrothermal reactions into aqueous rock stable for a long period of time to obtain solidification products insoluble to water and with an extremely low leaching rate. (Ikeda, J.)

Alternative waste management concept for medium and low level wastes by in-situ solidification

International Nuclear Information System (INIS)

Kraemer, R.

1982-01-01

Since 1976, a German R and D project has been carried out to find an alternative concept for the treatment and disposal of MLW and LLW arising mainly in the planned German reprocessing plant and other nuclear facilities (LWR, fuel fabrication, R and D establishments). The main feature of this concept is an in-situ solidification of preconditioned waste granules in large salt caverns located in the deep geological underground, thus avoiding such non-radioactive ballast as lost concrete shielding and container material. (orig./RW)

Behaviour of Aluminosilicate Inorganic Matrix Sial During and After Solidification of Radioactive Sludge And Radioactive Spent Resins and Their Mixtures

International Nuclear Information System (INIS)

Lichvar, P.; Rozloznik, M.; Sekely, S.

2013-01-01

SIAL matrix is the product of inorganic compounds by polycondensation. The inorganic SIAL matrix was subject of study from the atomic level to the level of technological application. Infrared spectroscopy measurement and X-ray analysis confirmed that SIAL matrix is mostly formed by amorphous phase. This satisfies the required criteria for disposal of immobilized radioactive waste of appropriate properties into surface repository (Slovak and Czech Republic). First of all, leaching of radioactive nuclide Cs-137 and compressive strength of the final product were found. Various remotely operated devices have been designed, manufactured and successfully used for the radioactive sludge and sludge/resins mixture solidification at the Nuclear power plant (NPP) A-1, V-2 in Jaslovske Bohunice, at Mochovce NPP (Slovak Republic), at Dukovany NPP and at Temelin NPP (Czech Republic). The SIAL matrix was approved for waste package for sludge/resins mixture by the Slovak and Czech Nuclear Regulatory Authorities. (author)

Behaviour of Aluminosilicate Inorganic Matrix Sial During and After Solidification of Radioactive Sludge And Radioactive Spent Resins and Their Mixtures

Energy Technology Data Exchange (ETDEWEB)

Lichvar, P.; Rozloznik, M.; Sekely, S. [Amec Nuclear Slovakia, 919 31 Jaslovske Bohunice (Slovakia)

2013-09-15

SIAL matrix is the product of inorganic compounds by polycondensation. The inorganic SIAL matrix was subject of study from the atomic level to the level of technological application. Infrared spectroscopy measurement and X-ray analysis confirmed that SIAL matrix is mostly formed by amorphous phase. This satisfies the required criteria for disposal of immobilized radioactive waste of appropriate properties into surface repository (Slovak and Czech Republic). First of all, leaching of radioactive nuclide Cs-137 and compressive strength of the final product were found. Various remotely operated devices have been designed, manufactured and successfully used for the radioactive sludge and sludge/resins mixture solidification at the Nuclear power plant (NPP) A-1, V-2 in Jaslovske Bohunice, at Mochovce NPP (Slovak Republic), at Dukovany NPP and at Temelin NPP (Czech Republic). The SIAL matrix was approved for waste package for sludge/resins mixture by the Slovak and Czech Nuclear Regulatory Authorities. (author)

Radioactive Wastes Cementation during Decommissioning Of Salaspils Research Reactor

International Nuclear Information System (INIS)

Abramenkova, G.; Klavins, M.; Abramenkovs, A.

2009-01-01

This paper deals with information on the radioactive wastes cementation technology for decommissioning of Salaspils Research Reactor (SRR). Dismantled radioactive materials were cemented in concrete containers using tritiated water-cement mortar. The laboratory tests system was developed to meet the waste acceptance criteria for disposal of containers with cemented radioactive wastes in near-surface repository 'Radons'. The viscosity of water-cement mortar, mechanical tests of solidified mortar's samples, change of temperature of the samples during solidification time and leakage of Cs-137 and T-3 radionuclides was studied for different water-cement compositions with different additives. The pH and electro conductivity of the solutions during leakage tests were controlled. It was shown, that water/cement ratio significantly influences on water-cement mortar's viscosity and solidified samples mechanical stability. Increasing of water ratio from 0.45 up to 0.62 decreases water-cement mortar's viscosity from 1100 mPas up to 90 mPas and decreases mechanical stability of water-cement samples from 23 N/mm 2 to the 12 N/mm 2 . The role of additives - fly ash and Penetron admix in reduction of solidification temperature is discussed. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature from 81 deg. C up to 62 deg. C. The optimal interval of water ratio in cement mortar is discussed. Tritium and Cs-137 leakage tests show, that radionuclides release curves has a complicate structure. The possible radionuclides release mechanisms are discussed. Experimental results indicated that addition of fly ash result in facilitation of tritium and cesium leakage in water phase. Further directions of investigations are drafted. (authors)

A conceptual chemical solidification/stabilization system to remediate radioactive raffinate sludge

International Nuclear Information System (INIS)

Carpenter, D.J.; Ansted, J.P.; Foldyna, J.T.

1994-01-01

Past operations at the U.S. Department of Energy's (DOE) Weldon Spring, Missouri, Superfund Site included the manufacture of nitroaromatic-based munitions and the production of uranium and thorium metal from ore concentrates. These operations generated a large quantity of diverse contaminated waste media including raffinate sludge, soil, sediment, and building debris. These various waste media are contaminated with varying amounts of radionuclides nitroaromatics, metals, metalloids, non-metals, polychlorinated biphenyls (PCBs) and asbestos. The volumes and diversity of contaminants and waste media pose significant challenges in identifying applicable remedial technologies, particularly for the excavation and treatment of the water-rich raffinate sludge. This paper presents the results of comprehensive efforts to develop a conceptual chemical solidification/stabilization (CSS) system to treat a variety of waste media. The emphasis of this paper is the treatment of a water-rich refractory raffinate sludge and site contaminated soils both radioactive and nonradioactive. The conceptual system design includes raffinate sludge excavation, dewatering, and CSS processing (reagent selection and formulation, reagent and waste storage and metering, and product mixing). Many innovations were incorporated into the design, producing a system that can process the various waste types. Additionally, the radioactive and hazardous constituents are sufficiently immobilized to allow the secured disposal in a waste cell of the treated product. The conceptual CSS system can also produce a variety of treated product types, ranging from a monolithic form to a compactible soil-like medium. The advantages of this system flexibility are also presented

Solidification of radioactive wastes by bituminization technology. II

International Nuclear Information System (INIS)

Stuchlik, S.; Brzobohaty, J.

1985-01-01

The bituminization line consists of a rotor film evaporator, a condenser, reserve tanks of bitumen emulsion and model concentrates, a roller train, a vapour concentrate tank, a control console, a bitumen emulsion pump, a model concentrates pump, and a tank for model concentrate preparation. Anion-active emulsion Silembit S-60 produced by Paramo Pardubice was chosen as the bitumen emulsion. A block diagram is given of the experimental bituminization line and its processing is described of model nonactive concentrates whose composition corresponds to that of actual wastes from the V-1 nuclear power plant. Successful tests of the line showed that it could be used for the disposal of radioactive wastes. (E.S.)

Mobile calcination and cementation unit for solidification of concentrated radioactive wastes

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Skaba, V.; Skvarenina, R.; Ditl, P.

1985-01-01

Mobile experimental unit MESA-1 was developed and manufactured for processing radioactive concentrates by direct cementation. The unit is mainly designed for processing low-level liquid wastes from nuclear power plants and other nuclear installations, in which the level of radioactivity does not exceed 10 10 Bq/m 3 , the salt content of liquid solutions does not exceed 500 kg/m 3 and the maximum amount of boric acid is 130 kg/m 3 . The equipment is built into three modules which may be assembled and dismantled in a short time and transported separately. The unit without the calciner module was tested in non-radioactive mode and in operation with actual radioactive wastes from the V-1 nuclear power plant. The course and results of the tests are described in detail. All project design values were achieved, a total of 18 dm 3 model solutions were processed and 1 m 3 of actual wastes with a salt content of 450 kg/m 3 . The test showed that with regard to the radiation level reached it will be necessary in the process of calcination to increase the shielding of certain exposed points. The calciner module is being assembled for completion. (Z.M.)

Low-level radioactive wastes in subsurface soils

International Nuclear Information System (INIS)

Francis, A.J.

1985-01-01

Low-level radioactive wastes will continue to be buried in shallow-land waste disposal sites. Several of the burial sites have been closed because of the problems that developed as a result of poor site characteristics, types of waste buried, and a number of other environmental factors. Some of the problems encountered can be traced to the activities of microorganisms. These include microbial degradation of waste forms resulting in trench cover subsidence, production of radioactive gases, and production of microbial metabolites capable of complexation, solubilization, and bioaccumulation of radionuclides. Improvements in disposal technology are being developed to minimize these problems. These include waste segregation, waste pretreatment, incineration, and solidification. Microorganisms are also known to enhance and inhibit the movement of metals. Little is known about the role of autotrophic microbial transformations of radionuclides. Such microbial processes may be significant in light of improved disposal procedures, which may result in reductions in the organic content of the waste disposed of at shallow-land sites. 102 references, 5 figures, 19 tables

Study on the High Volume Reduction of Radioactive Wastes

Energy Technology Data Exchange (ETDEWEB)

Kim, Ki Hong; Sik, Kang Il; Seok, Hong Dae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ho, Jeon Gil [RADIN Co. Ltd., Daejeon (Korea, Republic of)

2013-10-15

The solidification of radioactive wastes by the mixing method always increases their volume due to the limitation of incorporation ratio (waste/solidification agent). But if the powdered wastes can be compacted as the high density pellets and also the pellets can be filled up in a waste drum as much as possible while solidifying them with a very sticky solidification agent including a void formed in the filling step of pellets, it might be more desirable to reduce the waste volume as compared with the mixing method. So in this study, we designed and manufactured a high volume reduction machine which has the special size and shape of a pellet pocket, which the pellets can be extracted from easily and filled up in a large amount in drum, a pressurizing device to press 2 rolls, and the uniform feeding device of powder to the roll tyre. Some operational parameters which affect the formation of pellets from a powder were investigated, and then the volume reduction of a powder was evaluated. The briquetting machine, popular in general industry, was modified to apply for the volume reduction of the powered radioactive wastes (dried concentrate, sludge, spent ion-exchange resin, ash, depleted uranium powder, and etc.). In this developed high volume reduction machine, the capacity was 25 ∼ 62.5 kg/h at the optimum conditions, and the estimated volume reduction was about 2.95 (2.74/0.93) on the basis of between a powder (bulk density = 0.93 g/cm{sup 3}) and the pellet (2.74 g/cm{sup 3}). But on the basis of 200L drum, the calculated volume reduction was about 1.34 in consideration of a void volume originated in the filling step of the pellets.

Solidification of metal oxide from electrokinetic-electrodialytic decontamination

Energy Technology Data Exchange (ETDEWEB)

Koo, Daeseo; Park, Uk-Ryang; Kim, Gye-Nam; Kim, Seung-Soo; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Electrokinectic-electrodialytic decontamination technology reduced 80% of the concentration of the uranium soil waste to below the concentration of self-disposal. After conducting electrokinectic-electrodialytic decontamination, more than 10% of the remainder of radioactive waste from the cathodes of electrokinectic-electrodialytic equipment were produced. To dispose of such waste, it is necessary to solidify second radioactive waste owing to the requirements of radioactive waste from public corporations. In this study, a solidification experiment was carried out using a polymer. At first, a sampling of second radioactive waste was conducted. Then, second radioactive waste and a polymer were mixed. Third, the solidified state between the second radioactive waste and polymer was checked. In our next study, an experiment for the requirements of a public radioactive waste corporation will be conducted.

Main approaches to solving the problems of radioactive waste processing and disposal

International Nuclear Information System (INIS)

Tarasov, V.M.; Syrkus, M.N.

1989-01-01

papers devoted to the problems of processing and disposal of radioactive wastes, formed during nuclear facility operation and after decommissioning are reviewed. Techniques for gaseous and liquid waste solidification, as well as solid waste processing by mechanical fragmentation and combustion are considered. Possibilities of radioactive waste disposal in cosmic space, their burial at the bed of seas ansd oceans, in geological storages are discussed. Special attention disposal. The conclusion is made that today there are no any uniform way for radiactive waste disposal and standard technical means for its realization. Solution of the problems considered should be of a complex character and it is carried out within international research programs

Concentration and solidification of liquid radioactive wastes. Laboratory studies

International Nuclear Information System (INIS)

Nuche Vazquez, F.; Lora Soria, F. de

1969-01-01

Bench scale runs on concentration of intermediate level radioactive wastes, and incorporation of the concentrates in asphalt, are described. The feasibility of the process has been demonstrated, with a maximum incorporation of 60 percent of salts into the asphaltic matrix and a volume reduction factor of 10. (Author) 14 refs

Low level radioactive waste disposal/treatment technology overview: Savannah River site

International Nuclear Information System (INIS)

Sturm, H.F. Jr.

1987-01-01

The Savannah River Site will begin operation of several low-level waste disposal/treatment facilities during the next five years, including a new low-level solid waste disposal facility, a low-level liquid effluent treatment facility, and a low-level liquid waste solidification process. Closure of a radioactive hazardous waste burial ground will also be completed. Technical efforts directed toward waste volume reduction include compaction, incineration, waste avoidance, and clean waste segregation. This paper summarizes new technology being developed and implemented. 11 refs., 1 fig

Method of storing solidification products

International Nuclear Information System (INIS)

Tani, Yutaro.

1985-01-01

Purpose: To enable to efficiently and satisfactorily cool and store solidification products of liquid wastes generated from the reactor spent fuel reprocessing process by a simple facility. Method: Liquid wastes generated from the reactor spent fuel reprocessing process are caused to flow from the upper opening to the inside of a spherical canistor. The opening of the spherical canistor is welded with a lid by a remote control and the liquid wastes are tightly sealed within the spherical canistor as glass solidification products. Spherical canistors having the solidification products tightly sealed therein are sent into and stored in a hopper by the remote control. Further, a blower is driven upon storing to suck cooling air from the cooling air intake port to the inside of the hopper to absorb the decay heat of radioactive materials in the solidification products and the air is discharged from the duct and through the stack to the atmosphere. (Kawakami, Y.)

Solidification of high level liquid waste (HLLW) into ceramics by sintering process

International Nuclear Information System (INIS)

Masuda, Sumio; Oguino, Naohiko; Tsunoda, Naomi; O-oka, Kazuo; Ohta, Takao.

1979-01-01

One of the alternatives to vitrified solid which is acceptable and well characterized for storing radioactive HLLW with desirable long-term stability is ceramics. On the other hand, the solidification process of highly radioactive wastes should be simple and suitable for continuous production. On the above described basis, the authors have made preliminary study on the production of sintered ceramics by the addition of several oxides to HLLW. The simulated waste and additive oxides were pressed in a mold to make the preforms of 50 mm diameter and 10 to 15 mm thick. The preforms were then normally sintered at temperature from 1000 to 1400 deg C for 2 to 4 hours. The characterization of the sintered solids revealed the following facts. (1) X-ray diffraction analysis showed that the expected crystals were formed by normal-sintering as well as by hot-pressing. (2) The bulk density of the ceramics by normal-sintering was around 90 to 95% of the assumed theoretical values. (3) The leach-rate of the solids was affected by the bulk density. (4) Other properties of the solids, such as thermal expansion or thermal conductivity, are dominantly determined by those of main crystals in the solids. Sintering process is generally simple and productive as far as normal sintering is concerned. However, hot-pressing is an intermittent and time consuming process. From this fact, the authors intended to adopt the normal sintering process for the ceramic solidification of high level liquid wastes. (Wakatsuki, Y.)

Polymer solidification national program

International Nuclear Information System (INIS)

Kalb, P.D.; Colombo, P.

1993-04-01

Brookhaven National Laboratory (BNL) has developed several new and innovative polymer processes for the solidification of low-level radioactive, hazardous and mixed wastes streams. Polyethylene and modified sulfur cement solidification technologies have undergone steady, gradual development at BNL over the past nine years. During this time they have progressed through each of the stages necessary for logical technology maturation: from process conception, parameter optimization, waste form testing, evaluation of long-term durability, economic analysis, and scale-up feasibility. This technology development represents a significant investment which can potentially provide DOE with both short- and long-term savings

Radioactive waste management in the USSR: A review of unclassified sources

International Nuclear Information System (INIS)

Bradley, D.J.

1991-03-01

The Soviet Union does not currently have an overall radioactive waste management program or national laws that define objectives, procedures, and standards, although such a law is being developed, according to the Soviets. Occupational health and safety does not appear to receive major attention as it does in Western nations. In addition, construction practices that would be considered marginal in Western facilities show up in Soviet nuclear power and waste management operations. The issues involved with radioactive waste management and environmental restoration are being investigated at several large Soviet institutes; however, there is little apparent interdisciplinary integration between them, or interaction with the USSR Academy of Sciences. It is expected that a consensus on technical solutions will be achieved, but it may be slow in coming, especially for final disposal of high-level radioactive wastes and environmental restoration of contaminated areas. Meanwhile, many treatment, solidification, and disposal options for radioactive waste management are being investigated by the Soviets

Radioactive waste treatment at the Boris Kidric Institute of nuclear sciences

International Nuclear Information System (INIS)

Vukovic, Z.

1989-01-01

The results of many years work on the problems of treatment and interim storage of radioactive waste at the Boris Kidric Institute of nuclear sciences are presented. The main R/D work based on chemical treatment, solidification and pressing is described (author)

Radioactive waste treatment at the Boris Kidric Institute of nuclear sciences

Energy Technology Data Exchange (ETDEWEB)

Vukovic, Z [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1989-07-01

The results of many years work on the problems of treatment and interim storage of radioactive waste at the Boris Kidric Institute of nuclear sciences are presented. The main R/D work based on chemical treatment, solidification and pressing is described (author)

Radioactive waste management at the Hanford Reservation

International Nuclear Information System (INIS)

Anon.

1979-01-01

During some 30 years of plutonium production, the Hanford Reservation has accumulated large quantities of low- and high-level radioactive wastes. The high-level wastes have been stored in underground tanks, and the low-level wastes have been percolated into the soil. In recent years some programs for solidification and separation of the high-level wastes have been initiated. The Hanford waste-management system was studied by a panel of the Committee on Radioactive Waste Management of the National Academy of Sciences. The panel concluded that Hanford waste-management practices were adequate at present and for the immediate future but recommended increased research and development programs related to long-term isolation of the wastes. The panel also considered some alternatives for on-site disposal of the wastes. The Hanford Reservation was originally established for the production of plutonium for military purposes. During more than 30 years of operation, large volumes of high- and low-level radioactive wastes have been accumulated and contained at the site. The Management of these wastes has been the subject of controversy and criticism. To obtain a true technical evaluation of the Hanford waste situation, the Energy Research and Development Administration (now part of the Department of Energy) issued a contract to the National Academy of Sciences and the National Research Councilto conduct an independent review and evaluation of the Hanford waste-management practices and plans. A panel of the Committee on Radioactive Waste Management (CRWM) of the National Academy of Sciences conducted this study between the summer of 1976 and the summer of 1977. This article is a summary of the final report of that panel

Solidification of aqueous radioactive waste using insoluble compounds of magnesium oxide

International Nuclear Information System (INIS)

Carlson, J.E.

1986-01-01

A process is described for the treatment of radioactive waste which comprises: (a) first adding, under continuous agitation, a sufficient amount of a powdered magnesium oxide or magnesium hydroxide to an aqueous radioactive waste solution containing boric acid, the temperature of the water solution being 55-95 degrees C. to produce a magnesium borate derivative; (b) adding cement, under continuous agitation, to the magnesium borate derivative; and (c) then adding, under continuous agitation, after the cement has been dispersed, a sufficient amount of a compound selected from the group consisting of calcium oxide and calcium hydroxide to (b) to produce a gel matrix structure

Crystallization of sodium nitrate from radioactive waste

International Nuclear Information System (INIS)

Krapukhin, V.B.; Krasavina, E.P.; Pikaev, A.K.

1997-07-01

From the 1940s to the 1980s, the Institute of Physical Chemistry of the Russian Academy of Sciences (IPC/RAS) conducted research and development on processes to separate acetate and nitrate salts and acetic acid from radioactive wastes by crystallization. The research objective was to decrease waste volumes and produce the separated decontaminated materials for recycle. This report presents an account of the IPC/RAS experience in this field. Details on operating conditions, waste and product compositions, decontamination factors, and process equipment are described. The research and development was generally related to the management of intermediate-level radioactive wastes. The waste solutions resulted from recovery and processing of uranium, plutonium, and other products from irradiated nuclear fuel, neutralization of nuclear process solutions after extractant recovery, regeneration of process nitric acid, equipment decontamination, and other radiochemical processes. Waste components include nitric acid, metal nitrate and acetate salts, organic impurities, and surfactants. Waste management operations generally consist of two stages: volume reduction and processing of the concentrates for storage, solidification, and disposal. Filtration, coprecipitation, coagulation, evaporation, and sorption were used to reduce waste volume. 28 figs., 40 tabs

Glasses for the solidification of high-level radioactive waste: their behavior in the presence of water

International Nuclear Information System (INIS)

Grauer, R.

1983-02-01

Because of their amorphous structure, glasses are particularly suitable for the solidification of the mixture of high-level radioactive wastes resulting from reactor fuel reprocessing: they are not sensitive to variations in the compositions of waste oxides and are resistant to the damaging effects of radiation. The borosilicate glasses used for this purpose have been investigated for about 25 years, and waste vitrification techniques have been tested on a commercial scale. In view of possible accidents in a final waste repository, the chemical resistance of this type of glasses to attack by groundwaters is of special interest. The present report deals with the corrosion behaviour of glasses and discusses the most significant controlling parameters. The dissolution rates needed for safety analysis must be determined in relatively short-term experiments. Since the results can depend strongly on the type of test procedures used, a critical assessment of these techniques is necessary. Experimental results are illustrated by means of selected examples. Particular emphasis is placed upon the effects of increased temperatures and of nuclear radiation. The models which have been proposed for the estimation of the long-term behavior of vitrified waste are not yet fully complete and require improvement. Furthermore, the actual dissolution rates which are used in such models should be revised: to be desired are values which take into account the actual environmental conditions at the storage site. It should be noted, however, that even with current conservative input data on corrosion rates, a lifetime on the order of 10 5 years can be expected for the glass blocks to be deposited. The report concludes with recommendations fo further investigations

Immobilization of low and intermediate level of organic radioactive wastes in cement matrices

International Nuclear Information System (INIS)

Eskander, S.B.; Abdel Aziz, S.M.; El-Didamony, H.; Sayed, M.I.

2011-01-01

Highlights: → Solidification/stabilization of liquid scintillation waste. → Resistance to frost attack. → Retarding effect of scintillator waste was overcome by adding clay. → Evaluation of the suitability of cement-clay composite to solidify and stabilize scintillation waste. - Abstract: The adequacy of cement-clay composite, for solidification/stabilization of organic radioactive spent liquid scintillator wastes and its resistance to frost attack were determined by a freezing/thawing (F/T) test. Frost resistance is assessed for the candidate cement-clay composite after 75 cycles of freezing and thawing by evaluating their mass durability index, compressive strength, apparent porosity, volume of open pores, water absorption, and bulk density. Infrared (IR), X-ray diffraction (XRD), differential thermal analysis (DTA), thermal gravimetric analysis (TGA) and scanning electron microscopy (SEM) were performed for the final waste form (FWF) before and after the F/T treatment to follow the changes that may take place in its microstructure during the hydration regime. The results were obtained indicate that the candidate composite exhibits acceptable resistance to freeze/thaw treatment and has adequate suitability to solidify and stabilize organic radioactive spent liquid scintillator wastes even at very exaggerating conditions (-50 ° C and +60 ° C ).

Solidification of ion exchange resin wastes

International Nuclear Information System (INIS)

1982-08-01

Solidification media investigated included portland type I, portland type III and high alumina cements, a proprietary gypsum-based polymer modified cement, and a vinyl ester-styrene thermosetting plastic. Samples formulated with hydraulic cement were analyzed to investigate the effects of resin type, resin loading, waste-to-cement ratio, and water-to-cement ratio. The solidification of cation resin wastes with portland cement was characterized by excessive swelling and cracking of waste forms, both after curing and during immersion testing. Mixed bed resin waste formulations were limited by their cation component. Additives to improve the mechanical properties of portland cement-ion exchange resin waste forms were evaluated. High alumina cement formulations dislayed a resistance to deterioration of mechanical integrity during immersion testing, thus providing a significant advantage over portland cements for the solidification of resin wastes. Properties of cement-ion exchange resin waste forms were examined. An experiment was conducted to study the leachability of 137 Cs, 85 Sr, and 60 Co from resins modified in portland type III and high alumina cements. The cumulative 137 Cs fraction release was at least an order of magnitude greater than that of either 85 Sr or 60 Co. Release rates of 137 Cs in high alumina cement were greater than those in portland III cement by a factor of two.Compressive strength and leach testing were conducted for resin wastes solidified with polymer-modified gypsum based cement. 137 Cs, 85 Sr, and 60 Co fraction releases were about one, two and three orders of magnitude higher, respectively, than in equivalent portland type III cement formulations. As much as 28.6 wt % dry ion exchange resin was successfully solidified using vinyl ester-styrene compared with a maximum of 25 wt % in both portland and gypsum-based cement

Cementification for radioactive waste including high-concentration sodium sulfate and high-concentration radioactive nuclide

International Nuclear Information System (INIS)

Miyamoto, Shinya; Sato, Tatsuaki; Sasoh, Michitaka; Sakurai, Jiro; Takada, Takao

2005-01-01

For the cementification of radioactive waste that has large concentrations of sodium sulfate and radioactive nuclide, a way of fixation for sulfate ion was studied comprising the pH control of water in contact with the cement solid, and the removal of the excess water from the cement matrix to prevent hydrogen gas generation with radiolysis. It was confirmed that the sulfate ion concentration in the contacted water with the cement solid is decreased with the formation of ettringite or barium sulfate before solidification, the pH value of the pore water in the cement solid can control less than 12.5 by the application of zeolite and a low-alkali cement such as alumina cement or fly ash mixed cement, and removal of the excess water from the cement matrix by heating is possible with aggregate addition. Consequently, radioactive waste including high-concentration sodium sulfate and high-concentration radioactive nuclide can be solidified with cementitious materials. (author)

Method for solidification of radioactive iodine-containing solid wastes

International Nuclear Information System (INIS)

Ozawa, Yoshihiro; Funabashi, Kiyomi; Uetake, Naoto.

1987-01-01

Purpose: To process radioactive iodine containing solid wastes as non-leaching solidified wastes with no risk of iodine release. Method: It has been known for the thermal stability of CuI, PbI 2 or adsorbents containing the same that they do not release iodine in an inert gas atmosphere or in a reducing atmosphere at a temperature lower than 480 deg C. In view of the above, adsorbents containing iodine in the chemical form of CuI or PbI 2 , or CuI or powdery PbI 2 per se are sealed and solidified into low melting glass at a temperature of lower than 480 deg C at which no iodine release occurs in a non-oxidative atmosphere. Since the products are vitrified wastes, they scarcely show leaching property and are excellent in durability and stability. (Takahashi, M.)

Treatment and disposal of radioactive wastes from the viewpoint of the NUCLEX 78

Energy Technology Data Exchange (ETDEWEB)

Koerner, W [Staatliches Amt fuer Atomsicherheit und Strahlenschutz, Berlin (German Democratic Republic)

1980-02-01

The results and consequences of the NUCLEX 78 are considered in form of a progress report on treatment and disposal of radioactive wastes from the nuclear fuel cycle. Investigations performed in the USA, Western Europe, and Japan are concerned with rationalization of the treatment processes for low-level and intermediate-level radioactive wastes and with the development of industrial methods of high-level waste solidification. In the field of ultimate storage, utilization of stable rock layers in the deep underground - especially of salt rocks - is evaluated to be the only available method of long-term isolation of high-level radioactive wastes and wastes containing long-lived alpha emitters. After technical and economical as well as safety works will have been concluded, commissioning of repositories in the underground is to be expected in the mid nineties.

Treatment and conditioning of low-level radioactive waste in Belgium: initial operating results of the Cilva facility

International Nuclear Information System (INIS)

Monsch, O.; Renard, C.; Deckers, J.; Luycx, P.

1995-01-01

The Belgian National Radioactive Waste and Enriched Fissile Material Agency (ONDRAF), which is responsible for the management of all radioactive waste in Belgium, recently decided to commission the CILVA facility. Operation of this facility, which comprises a number of units for the treatment of low-level radwaste, has been contracted to ONDRAF's Belgoprocess subsidiary based at the Dessel site. A consortium comprising SGN and Fabricom was in charge of building the CILVA facility's waste preparation and conditioning (concrete solidification) units. The concrete solidification processes, which were devised and developed by SGN, have been qualified to secure ONDRAF certification of the process and the facility. This enabled active commissioning of the waste conditioning unit in mid-August 1994. Active commissioning of the waste preparation unit was carried out in several stages up to the beginning of 1995 in accordance with operating requirements. Initial operating results of the two units are presented. (author)

Operation technology of the ventilation system of the radioactive waste treatment facility(II) - Design and operation note

Energy Technology Data Exchange (ETDEWEB)

Lee, K. M.; Lee, B. C.; Bae, S. M. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

As the radioactive waste treatment work, such as compaction and/or solidification of wastes, are done directly by the workers in the Radioactive Waste Treatment Facility, the reasonable design and operation of the ventilation system is essential. In this report, the design criteria and specification of the ventilation equipment, system operation method are described for the effective design and operation of ventilation system in the radioactive waste treatment facility. And the anti-vibration work which was done in the Radioactive Waste Treatment Facility in KAERI to reduce the effect of vibration due to the continuous operation of big rotational equipment, the intake fans and the exhaust fans, are described in the report. 11 refs., 10 figs., 12 tabs. (Author)

Characterization of surrogate radioactive cemented waste: a laboratory study

International Nuclear Information System (INIS)

Fiset, J.F.; Lastra, R.; Bilodeau, A.; Bouzoubaa

2011-01-01

Portland cement is commonly used to stabilize intermediate and low level of radioactive wastes. The stabilization/solidification process needs to be well understood as waste constituents can retard or activate cement hydration. The objectives of this project were to prepare surrogate radioactive cemented waste (SRCW), develop a comminution strategy for SRCW, determine its chemical characteristics, and develop processes for long term storage. This paper emphasizes on the characterization of surrogate radioactive cemented waste. The SRCW produced showed a high degree of heterogeneity mainly due to the method used to add the solution to the host cement. Heavy metals such as uranium and mercury were not distributed uniformly in the pail. Mineralogical characterization (SEM, EDS) showed that uranium is located around the rims of hydrated cement particles. In the SRCW, uranium occurs possibly in the form of a hydrated calcium uranate.The SEM-EDS results also suggest that mercury occurs mainly in the form of HgO although some metallic mercury may be also present as a result of partial decomposition of the HgO. (author)

Developing technologies for conditioning the liquid organic radioactive wastes from Cernavoda NPP

International Nuclear Information System (INIS)

Deneanu, N.; Popescu, I. V.; Teoreanu, I.

2004-01-01

The Institute for Nuclear Research (INR)-Pitesti has developed technologies for conditioning liquid organic radioactive wastes (oils, miscellaneous solvent and liquid scintillation cocktail) for Cernavoda NPP. This paper describes the new and viable solidification technology to convert liquid organic radioactive wastes into a stable monolithic form, which minimizes the probability to release tritium in the environment during interim storage, transportation and final disposal. These are normally LLW containing only relatively small quantities of beta/gamma emitting radionuclides and variable amounts of tritium with activity below E+08Bq/l. The INR research staff in the radwaste area developed treatment/conditioning techniques and also designed and tested the containers for the final disposal, following the approach in the management of radwaste related to the nuclear fuel cycle. Thus, the INR focused this type of activity on treating and conditioning the wastes generated at Cernavoda Nuclear Power Plant consisting of lubricants from primary fuelling machines and turbine, the miscellaneous solvent from decontamination operation and the liquid scintillation cocktail used in radiochemical analysis. Laboratory studies on cementation of liquid organic radioactive wastes have been undertaken at INR Pitesti. One simple system, similar to a conventional cement solidification unit, can treat radioactive liquid wastes, which are the major components of low- and medium-level radioactive wastes generated by a Nuclear Power Plant. It was proved that the solidified waste could meet the Waste Acceptance Criteria of the disposal site, in this case Baita-Bihor National Repository, as follows: - The wastes are deposited in type A packages; - The maximum expected quantities of this waste stream that will be produced in the future are 50 drums per year. The maximum specific tritium activity per drum is 10 9 Bq/m 3 ; - Compressive strengths of the samples should be greater than 50 MPa

Solidification of ion exchange resin wastes in hydraulic cement

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Kalb, P.; Fuhrmann, M.; Colombo, P.

1982-01-01

Work has been conducted to investigate the solidification of ion exchange resin wastes with portland cements. These efforts have been directed toward the development of acceptable formulations for the solidification of ion exchange resin wastes and the characterization of the resultant waste forms. This paper describes formulation development work and defines acceptable formulations in terms of ternary phase compositional diagrams. The effects of cement type, resin type, resin loading, waste/cement ratio and water/cement ratio are described. The leachability of unsolidified and solidified resin waste forms and its relationship to full-scale waste form behavior is discussed. Gamma irradiation was found to improve waste form integrity, apparently as a result of increased resin crosslinking. Modifications to improve waste form integrity are described. 3 tables

Heat transfer enhanced microwave process for stabilization of liquid radioactive waste slurry. Final report

International Nuclear Information System (INIS)

White, T.L.

1995-01-01

The objectve of this CRADA is to combine a polymer process for encapsulation of liquid radioactive waste slurry developed by Monolith Technology, Inc. (MTI), with an in-drum microwave process for drying radioactive wastes developed by Oak Ridge National Laboratory (ORNL), for the purpose of achieving a fast, cost-effectve commercial process for solidification of liquid radioactive waste slurry. Tests performed so far show a four-fold increase in process throughput due to the direct microwave heating of the polymer/slurry mixture, compared to conventional edge-heating of the mixer. We measured a steady-state throughput of 33 ml/min for 1.4 kW of absorbed microwave power. The final waste form is a solid monolith with no free liquids and no free particulates

Plastic solidification system at Hamaoka Nuclear Power Station

International Nuclear Information System (INIS)

Okajima, Hiroyuki; Iokibe, Hiroyuki; Tsukiyama, Shigeru; Suzuki, Michio; Yamaguchi, Masato

1987-01-01

In Unit 1 and 2 of the Hamaoka Nuclear Power Station, radioactive waste was previously solidified in cement. By this method, the quantity of waste thus treated is relatively small, resulting in large number of the solidified drums. In order to solve this problem, the solidification facility using a thermosetting resin was employed, which is in operation since January 1986 for Unit 1, 2 and 3. As compared with the cement solidification, the solidified volume of concentrated liquid is about 1/12 and of spent-resin slurry is about 1/4 in plastic solidification. The following are described: course leading to the employment, the plastic solidification facility, features of the facility, operation results so far with the facility, etc. (Mori, K.)

Application of concrete to the treatment and disposal of radioactive waste in Japan

International Nuclear Information System (INIS)

Maki, Yasuro; Ohnuma, Hiroshi

1992-01-01

The paper presents the present state of application of concrete to treatment, storage and disposal of low level radioactive waste in Japan. In the 2nd section, the electric power supply and the kinds and volumes of radioactive waste from nuclear power plants in Japan are described. In the 3rd section, the applications of concrete to the treatment of radioactive waste are described. These are solidification with cement and containers made by various mortars and concretes. The application of concrete to disposal structures are presented in the 4th section; these are research on the durabity of concrete under disposal site condition, research on the filling the concrete pit with 200 l drum packed cement solidified wastes by prepacked concreting methods, and so on. And this section describes also the outlines of the low level radioactive disposal system at the Rokkasho site. (orig./DG)

Ordinary Portland Cement matrix for solidification of cellulosic protective clothes hazardous wastes

International Nuclear Information System (INIS)

Shatta, H.A.; Saleh, H.M.

2006-01-01

The used cellulosic protective clothes constitutes considerable fraction of the hazardous and radioactive wastes accumulated during the practical daily life. The direct solidification of these wastes with ordinary Portland cement resulted in waste forms having undesired characters, therefore, it is recommended to immobilize the secondary waste solutions coming from the oxidative degradation of the used protective clothes waste simulates rather than direct imbedding. IR analyses, X-ray diffraction and thermal characteristics for products of both direct encapsulation of the waste and the cementation of its degradation products were performed to evaluate the properties of the final waste cemented form before their disposal. Based on the results reached from X-ray diffraction, IR spectrograms and thermal analyses reports, it could be stated that no detectable changes in hydration and curing coarse of ordinary Portland cement when mixing the residual secondary waste solution resulting from the oxidative degradation of the used protective clothes waste simulate compared with mixing cement with water and in reverse with imbedding the unprocessed waste in cement matrix

Immobilization of low and intermediate level of organic radioactive wastes in cement matrices

Energy Technology Data Exchange (ETDEWEB)

Eskander, S.B. [Radioisotopes Department, Atomic Energy Authority, Dokki, Cairo (Egypt); Abdel Aziz, S.M. [Middle Eastern Regional Radioisotope Centre for the Arab Countries, Dokki, Cairo (Egypt); El-Didamony, H. [Faculty of Science, Zagazig University, Zagazig, El-Sharkia (Egypt); Sayed, M.I., E-mail: mois_161272@yahoo.com [Middle Eastern Regional Radioisotope Centre for the Arab Countries, Dokki, Cairo (Egypt)

2011-06-15

Highlights: {yields} Solidification/stabilization of liquid scintillation waste. {yields} Resistance to frost attack. {yields} Retarding effect of scintillator waste was overcome by adding clay. {yields} Evaluation of the suitability of cement-clay composite to solidify and stabilize scintillation waste. - Abstract: The adequacy of cement-clay composite, for solidification/stabilization of organic radioactive spent liquid scintillator wastes and its resistance to frost attack were determined by a freezing/thawing (F/T) test. Frost resistance is assessed for the candidate cement-clay composite after 75 cycles of freezing and thawing by evaluating their mass durability index, compressive strength, apparent porosity, volume of open pores, water absorption, and bulk density. Infrared (IR), X-ray diffraction (XRD), differential thermal analysis (DTA), thermal gravimetric analysis (TGA) and scanning electron microscopy (SEM) were performed for the final waste form (FWF) before and after the F/T treatment to follow the changes that may take place in its microstructure during the hydration regime. The results were obtained indicate that the candidate composite exhibits acceptable resistance to freeze/thaw treatment and has adequate suitability to solidify and stabilize organic radioactive spent liquid scintillator wastes even at very exaggerating conditions (-50{sup Degree-Sign }C and +60{sup Degree-Sign }C ).

Study on the conditions of bituminization of radioactive wastes and their influence on the stability of stored products

International Nuclear Information System (INIS)

Golinski, M.; Ksiazak, Z.

1975-05-01

Investigations carried out on a laboratory and semi-industrial scale showed that the Polish oxidised industrial bitumen P-60 was suitable for the solidification of liquid radioactive waste and particularly for non-concentrated post-precipitation sludges. The bitumen products were highly stable and were resistant to leaching by acids, salt solutions and water. Laboratory leach tests gave values similar to those obtained by others using different bitumen. By evaluating the sorption characteristics of the soil and the hydrogeological conditions existing at a proposed storage site, it was shown that the solidified wastes could be stored directly in the soil without further isolation from the soil water. Based on the liquid wastes arising from a nuclear power plant it has been shown that solidification of the wastes in bitumen will be cheaper than solidification of the same wastes using cement

Solidification of acidic liquid waste from 99Mo isotope production

International Nuclear Information System (INIS)

Parsons, G.J.

2001-01-01

Full text: The production of the radioisotope molybdenum-99 by the fission process began at ANSTO in the late 1960's. Molybdenum-99, with a half life of 66 hours, decays by beta emission to produce technetium-99m, a metastable isotope. Technetium-99m is the most widely used medical radioisotope due to its near ideal properties, particularly the radioactive half life of only 6 hours. ANSTO has been producing generators for around 30 years for distribution to hospitals and nuclear medicine centres. These generators produce technetium-99m for medical use by decay of the contained molybdenum-99. To produce molybdenum-99, uranium dioxide pellets enriched to 2.2% 235 U are irradiated in ANSTO's HIFAR reactor for about one week. The irradiated pellets are subsequently dissolved in nitric acid to allow the recovery of the molybdenum. An acidic intermediate level liquid waste results from this processing. A primary waste results from the raw leach solution (after removal of the molybdenum onto a packed alumina column) and a weaker secondary waste is produced from a series of column washing steps. The waste solution contains uranium, the majority of the other fission products and low levels of ammonia in a nitric acid solution. This liquid waste had been accumulating and stored in specially designed shielded tanks in a storage facility. A process has been developed at ANSTO to convert this intermediate level liquid waste into a crystalline solid form of considerably less volume and mass, for improved storage. The operation comprises three processing steps. The lower strength secondary waste solution first requires concentration, with the removal of water and some acid into a condensate. The condensate is chemically neutralised and treated through the conventional water treatment plant. Concentrated solution is then treated in a batch chemical process to reduce the low levels of ammonia to very low levels. The final evaporation process removes further water and acid and

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE.

Energy Technology Data Exchange (ETDEWEB)

ADAMA, J.W.; BOWERMAN, B.S.; KALB, P.D.

2002-10-01

The Environmental Protection Agency (EPA) is currently seeking to validate technologies that can directly treat radioactively contaminated high mercury (Hg) subcategory wastes without removing the mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needs additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 30 wt% dry sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes.

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

Energy Technology Data Exchange (ETDEWEB)

Adams, J. W.; Bowerman, B. S.; Kalb, P. D.

2002-02-25

The Environmental Protection Agency (EPA) is currently evaluating alternative treatment standards for radioactively contaminated high mercury (Hg) subcategory wastes, which do not require the removal of mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needed additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 46 wt% (30 wt% dry) sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide the EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes.

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

International Nuclear Information System (INIS)

ADAMA, J.W.; BOWERMAN, B.S.; KALB, P.D.

2002-01-01

The Environmental Protection Agency (EPA) is currently seeking to validate technologies that can directly treat radioactively contaminated high mercury (Hg) subcategory wastes without removing the mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needs additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 30 wt% dry sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes

SULFUR POLYMER STABILIZATION/SOLIDIFICATION (SPSS) TREATABILITY OF SIMULATED MIXED-WASTE MERCURY CONTAMINATED SLUDGE

International Nuclear Information System (INIS)

Adams, J. W.; Bowerman, B. S.; Kalb, P. D.

2002-01-01

The Environmental Protection Agency (EPA) is currently evaluating alternative treatment standards for radioactively contaminated high mercury (Hg) subcategory wastes, which do not require the removal of mercury from the waste. The Sulfur Polymer Stabilization/Solidification (SPSS) process developed at Brookhaven National Laboratory is one of several candidate technologies capable of successfully treating various Hg waste streams. To supplement previously supplied data on treatment of soils, EPA needed additional data concerning stabilization of high Hg subcategory waste sludges. To this end, a 5000 ppm sludge surrogate, containing approximately 50 wt% water, was successfully treated by pilot-scale SPSS processing. In two process runs, 85 and 95 wt% of water was recovered from the sludge during processing. At waste loadings of 46 wt% (30 wt% dry) sludge, the treated waste form had no detectable mercury (<10 ppb) in TCLP leachates. Data gathered from the demonstration of treatment of this sludge will provide the EPA with information to support revisions to current treatment requirements for high Hg subcategory wastes

Experience with radioactive waste incineration at Chalk River Nuclear Laboratories

International Nuclear Information System (INIS)

Le, V.T.; Beamer, N.V.; Buckley, L.P.

1988-06-01

Chalk River Nuclear Laboratories is a nuclear research centre operated by Atomic Energy of Canada Limited. A full-scale waste treatment centre has been constructed to process low- and intermediate-level radioactive wastes generated on-site. A batch-loaded, two-stage, starved-air incinerator for solid combustible waste is one of the processes installed in this facility. The incinerator has been operating since 1982. It has consistently reduced combustible wastes to an inert ash product, with an average volume reduction factor of about 150:1. The incinerator ash is stored in 200 L drums awaiting solidification in bitumen. The incinerator and a 50-ton hydraulic baler have provided treatment for a combined volume of about 1300 m 3 /a of solid low-level radioactive waste. This paper presents a review of the performance of the incinerator during its six years of operation. In addition to presenting operational experience, an assessment of the starved-air incineration technique will also be discussed

Characteristics of low-level radioactive waste disposed during 1987--1989

International Nuclear Information System (INIS)

Roles, G.W.

1990-12-01

This report presents the volume, activity, and radionuclide distributions in low-level radioactive waste (LLW) disposed during 1987 through 1989 at the commercial disposal facilities located near Barnwell, SC, Richland, WA, and Beatty, NV. The report has been entirely assembled from descriptions of waste provided in LLW shipment manifests. Individual radionuclide distributions are listed as a function of waste class, of general industry, and of waste stream. In addition, information is presented about disposal of wastes containing chelating agents, about use of solidification media, about the distribution of radiation levels at the surfaces of waste containers, and about the distribution of waste container sizes. Considerably more information is presented about waste disposed at the Richland and Beatty disposal facilities than at the Barnwell disposal facility

Development of cermets for high-level radioactive waste fixation

International Nuclear Information System (INIS)

Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

1979-01-01

A method is currently under development for the solidification and fixation of commercial and defense high-level radioactive wastes in the form of ceramic particles encapsulated by metal, i.e., a cermet. The chemical and physical processing techniques which have been developed and the properties of the resulting cermet bodies are described in this paper. These cermets have the advantages of high thermal conductivity and low leach rates

Radioactive waste management

International Nuclear Information System (INIS)

2013-01-01

This eighth chapter presents the radioactive wastes and waste disposal; classification of radioactive wastes; basis requests of the radioactive waste management; conditions for a radioactive waste disposal; registers and inventories; transport of radioactive wastes from a facility to another and the radioactive waste management plan

Techniques for the solidification of high-level wastes

International Nuclear Information System (INIS)

1977-01-01

The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

Liquid low-level waste (LLLW) solidification at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Schultz, R.M.; Monk, T.H.; duMont, S.P.; Helms, R.E.; Keigan, M.V.; Morris, M.I.

1987-01-01

In general, the presentation describes the disposal of liquid, low-level (radioactive) waste (LLLW) by the hydrofracture process at Oak Ridge National Laboratory until 1984, when it was shut down due to regulatory concerns and operational anomalies. As a result of this, about 400,000 gallons of concentrated LLLW and 50,000 gallons of transuranic waste-bearing sludges have accumulated in the active, double-contained tank system which is reaching its operational capacity. A major initiative to develop an alternative means of LLLW treatment and disposal was begun about two years ago. This presentation summarizes the implementation strategy of the most likely process options. The strategy is being developed in two phases; a near-term flowsheet and a long-term or reference flowsheet. First, reliable and fully demonstrated commercial, cement solidification systems are being assessed for execution of an initial 50,000 gallon campaign in 1988. Second, development is under way to determine viable sludge separation, LLLW decontamination and solidification alternatives. A flowsheet analysis and cost study is being conducted by a consultant to ensure proper consideration of process developments at other sites. It is estimated that, depending upon funding requirements, it could take up to six years to implement the reference flowsheet

Management of radioactive wastes - an overview of the Indian programme

International Nuclear Information System (INIS)

Thomas, K.T.; Sunder Rajan, N.S.; Balu, K.; Khan, A.A.

1977-01-01

An overview of the management of radioactive wastes with particular reference to the Indian Nuclear Programme is presented. The initial design philosophy of the radwaste management system is discussed in relation to accepting a calculated, minimum discharge of radioactivity to the environment. A brief report of the operational experience with the low and intermediate level radwaste systems is given. Factors that influence the review of the present philosophy for future adoption are presented. Some methods being developed for decreasing release of the radioactivity to the environment are discussed. Among techniques considered are solar evaporation, delay and decay of fission rare gases from power reactors and concentration and storage of Kr 85 from fuel reprocessing plants. Problems in the management of high level and alph-bearing wastes are discussed with particular reference to the nature of the waste generated and the policy under implementation for their management. The matrices, solidification processes, modes of interim storage and criteria for selection of site for ultimate dispensation of the solidified high level wastes in geological formations are described. An approach towards the solution of the probelm of management of alpha-bearing waste is also presented

Treatment of radioactive wastes containing plutonium

International Nuclear Information System (INIS)

Orlando, O.S.; Aparicio, G.; Greco, L.; Orosco, E.H.; Cassaniti, P.; Salguero, D.; Toubes, B.; Perez, A.E.; Menghini, J.E.; Esteban, A.; Adelfang, P.

1987-01-01

The radioactive wastes generated in the process of manufacture and control of experimental fuel rods of mixed oxides, (U,Pu)O 2 , require an specific treatment due to the plutonium content. The composition of liquid wastes, mostly arising from chemical checks, is variable. The salt content, the acidity, and the plutonium and uranium content are different, which makes necessary a chemical treatment before the inclusion in concrete. The solid waste, such as neoprene gloves, PVC sleeves, filter paper, disposable or broken laboratory material, etc. are also included in concrete. In this report the methods used to dispose of wastes at Alpha Facility are described. With regard to the liquid wastes, the glove box built to process them is detailed, as well as the applied chemical treatment, including neutralization, filtration and later solidification. As for the solid wastes, it is described the cementation method consisting in introducing them into an expanded metal matrix, of the basket type, that contains as a concentric drum of 200 liter capacity which is smaller than the matrix, and the filling with wet cement mortar. (Author)

U.S. Department of Energy's 'initiatives for proliferation prevention' program: solidification technologies for radioactive waste treatment in Russia - 16037

International Nuclear Information System (INIS)

Pokhitonov, Yuri; Kelley, Dennis

2009-01-01

Large amounts of liquid radioactive waste have existed in the U.S. and Russia since the 1950's as a result of the Cold War. Comprehensive action to treat and dispose of waste products has been lacking due to insufficient funding, ineffective technologies or no proven technologies, low priority by governments among others. Today the U.S. and Russian governments seek new, more reliable methods to treat liquid waste, in particular the legacy waste streams. A primary objective of waste generators and regulators is to find economical and proven technologies that can provide long-term stability for repository storage. In 2001, the V.G. Khlopin Radium Institute (Khlopin), St. Petersburg, Russia, and Pacific Nuclear Solutions (PNS), Indianapolis, Indiana, began extensive research and test programs to determine the validity of polymer technology for the absorption and immobilization of standard and complex waste streams. Over 60 liquid compositions have been tested including extensive irradiation tests to verify polymer stability and possible degradation. With conclusive scientific evidence of the polymer's effectiveness in treating liquid waste, both parties have decided to enter the Russian market and offer the solidification technology to nuclear sites for waste treatment and disposal. In conjunction with these efforts, the U.S. Department of Energy (DOE) will join Khlopin and PNS to explore opportunities for direct application of the polymers at predetermined sites and to conduct research for new product development. Under DOE's 'Initiatives for Proliferation Prevention' (IPP) program, funding will be provided to the Russian participants over a three year period to implement the program plan. This paper will present updated details of U.S. DOE's IPP program, the project structure and its objectives both short and long-term, polymer tests and applications for LLW, ILW and HLW, and new product development initiatives. (authors)

Evaluation of concrete as a matrix for solidification of Savannah River Plant waste

International Nuclear Information System (INIS)

Stone, J.A.

1977-06-01

The properties of concrete as a matrix for solidification of Savannah River Plant (SRP) high-level radioactive wastes were studied. In an experimental, laboratory-scale program, concrete specimens were prepared and evaluated with both simulated and actual SRP waste sludges. Properties of concrete were found adequate for fixation of SRP wastes. Procedures were developed for preparation of simulated sludges and concrete-sludge castings. Effects of cement type, simulated sludge type, sludge loading, and water content on concrete formulations were tested in a factorial experiment. Compressive strength, leachability of strontium and plutonium, thermal stability, and radiation stability were measured for each formulation. From these studies, high-alumina cement and a portland-pozzolanic cement were selected for additional tests. Incorporation of cesium-loaded zeolite into cement-sludge mixtures had no adverse effects on mechanical or chemical properties of waste forms. Effects of heating concrete-sludge castings were investigated; thermal conductivity and DTA-TGA-EGA data are reported. Formulations of actual SRP waste sludges in concrete were prepared and tested for compressive strength; for leachability of 90 Sr, 137 Cs, and alpha emitters; and for long-term thermal stability. The radioactive sludges were generally similar in behavior to simulated sludges in concrete. 37 tables, 34 figures

Method of solidifying liquid radioactive wastes

International Nuclear Information System (INIS)

Pekar, A.; Petrovic, J.; Timulak, J.

1987-01-01

Liquid radioactive waste containing boric acid salts is mixed with zeolite tuff and neutralized by lime. Power plant fly ash containing single-component or mixed Portland cement is then added to the mixture. Prior to packaging, anion-active bitumen emulsion or an aqueous emulsion of fatty acid salts and of free fatty acids insoluble in water can be added. Examples are given listing accurate proportions of the individual components. The advantage of the said solidification method is the use of easily available raw materials and improved values of extractability of the resulting product radionuclides. (E.S.)

Management of radioactive wastes from the nuclear fuel cycle

International Nuclear Information System (INIS)

1976-01-01

The increased emphasis in many countries on the development and utilization of nuclear power is leading to an expansion of all sectors of the nuclear fuel cycle, giving rise to important policy issues and radioactive-waste management requirements. Consequently, the IAEA and the Nuclear Energy Agency of OECD felt that it would be timely to review latest technology for the management of the radioactive wastes arising from nuclear fuel cycle facilities, to identify where important advances have been made, and to indicate those areas where further technological development is needed. Beginning in 1959, the IAEA, either by itself or jointly with OECD/NEA has held seven international symposia on the management of radioactive wastes. The last symposium, on the management of radioactive wastes from fuel reprocessing, was held jointly by the IAEA and OECD/NEA in Paris in November 1972. An objective of the 1976 symposium was to update the information presented at the previous symposia with the latest technological developments and thinking regarding the management and disposal of all categories of radioactive wastes. Consequently, although the scope of the symposium was rather broad, attention was focussed on operational experience and progress in unresolved areas of radioactive waste management. The programme dealt primarily with the solidification of liquid radioactive wastes and disposal of the products, especially the high-level fission products and actinide-containing waste from fuel reprocessing. Other topics covered policy and planning, treatment of hulls and solvent, management of plutonium-contaminated waste, and removal of gaseous radionuclides. The major topic of interest was the current state of the technology for the reduction and incorporation of the high-level radioactive liquid from fuel reprocessing into solid forms, such as calcines, glasses or ceramics, for safe interim storage and eventual disposal. The approaches to vitrification ranged from two stage

Production-scale LLW and RMW solidification system operational testing at Argonne National Laboratory-East (ANL-E)

International Nuclear Information System (INIS)

Wescott, J.; Wagh, A.; Singh, D.; Nelson, R.; No, H.

1997-01-01

Argonne National Laboratory-East (ANL-E) has begun production-scale testing of a low-level waste and radioactive mixed waste solidification system. This system will be used to treat low-level and mixed radioactive waste to meet land burial requirements. The system can use any of several types of solidification media, including a chemically bonded phosphate ceramic developed by ANL-E scientists. The final waste product will consist of a solidified mass in a standard 208-liter drum. The system uses commercial equipment and incorporates several unique process control features to ensure proper treatment. This paper will discuss the waste types requiring treatment, the system configuration, and operation results for these waste streams

Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

International Nuclear Information System (INIS)

Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

2007-01-01

Low-level liquid waste with relatively high levels of radioactivity is generated by the sulfuric acid elution treatment of spent ion exchange resin used in water purification systems of nuclear power plants. We studied cement-like solidification process for this type waste that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete, and this leads to cracking during repository storage. It is also preferable that the pH of pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Our solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste when the Ba/SO 4 molar ratio > 1. In the FLS procedure, silica fume and blast furnace slag are added to the liquid wastes containing Ba S O 4 precipitate. The CIS reaction yield is over 98% and the pH of pore water of the solid waste is 11.5 or less. Therefore, we think that our solidification process is one of the best methods for treating liquid waste that contains a high concentration of sodium sulfate. (author)

Experimental study on intermediate level radioactive waste processing

International Nuclear Information System (INIS)

Nagakura, Tadashi; Abe, Hirotoshi; Okazawa, Takao; Hattori, Seiichi; Maki, Yasuro

1977-01-01

In the disposal of intermediate level radioactive wastes, multilayer package will be adopted. The multilayer package consists of cement-solidified waste and a container such as a drum - can with concrete liner or a concrete container. So, on the waste to be cement-solidified in such container, experimental study was carried out as follows. (1) Cement-solidification method. (2) Mechanical behaviour of cement-solidified waste. The mechanical behaviour of the containers was studied by the finite element method and experiment, and the function of pressure-balancing valves was also studied. The following data on processing intermediate level radioactive wastes were obtained. (1) In the case of cement-solidified waste, the data to select the suitable solidifying material and the standard mixing proportion were determined. (2) The basic data concerning the uniaxial compressive strength of cement-solidified waste, the mechanical behaviour of cement-solidified waste packed in a drum under high hydrostatic pressure, the shock response of cement-solidified waste at the time of falling and so on were obtained. (3) The pressure-balancing valves worked at about 0.5 Kg/cm 2 pressure difference inside and outside a container, and the deformation of a drum cover was 10 to 13 mm. In case of the pressure difference less than 0,5 Kg/cm 2 , the valves shut, and water flow did occur. (auth.)

Long-term leach testing of solidified radioactive waste forms (International Standard Publication ISO 6961:1982)

International Nuclear Information System (INIS)

Stefanik, J.

2001-01-01

Processes are developed for the immobilization of radionuclides by solidification of radioactive wastes. The resulting solidification products are characterized by strong resistance to leaching aimed at low release rates of the radionuclides to the environment. To measure this resistance to leaching of the solidified materials: glass, glass-ceramics, bitumen, cement, concrete, plastics, a long-term leach test is presented. The long-term leach test is aimed at: a) the comparison of different kinds or compositions of solidified waste forms; b) the intercomparison between leach test results from different laboratories on one product; c) the intercomparison between leach test results on products from different processes

Development of high-level waste solidification technology 1

Energy Technology Data Exchange (ETDEWEB)

Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

1999-02-01

Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes

International Nuclear Information System (INIS)

1980-01-01

The objective of this study was to review the relative merits and potential of 15 (fifteen) alternative waste forms being considered for the solidification and disposal of radioactive wastes. The relative merits of 4 (four) alternative pre-solidification processing approaches were also assessed in this study. A Peer Review Panel composed of 8 (eight) scientists and engineers representing independent, non-DOE laboratories from industry, government, and universities and the disciplines of materials science, ceramics, glass, metallurgy, and geology conducted the review. A numerical rating of alternative waste forms was arrived at individually by the panel members taking into consideration 9 (nine) scientific and 9 (nine) engineering parameters affecting the long term performance and production of waste forms. At a meeting on May 9, 1980, a group ranking for the alternative forms was achieved by averaging the individual scores and discussing the available data base. Three final ranking lists comparing: (A) Present Scientific Merits or Least Risk for Use Today; and (B) Research Priority; and (C) Present and Potential Engineering Practicality were prepared by the Panel. Each waste form in the lists is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or (3) Bottom Rank. A discussion of the relative strengths and weaknesses of the alternative waste forms and recommendations for future program directions is presented in the body of the accompanying Peer Review Panel report

Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

Energy Technology Data Exchange (ETDEWEB)

Rashid, F.; Ghaffar, A. [Pakistan Inst. of Nuclear Science and Technology, Islamabad (Pakistan)

2011-07-01

The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

Sorption of radioscesium from liquid radioactive waste on clay and immobilization by baking the clay at elevated temperature

International Nuclear Information System (INIS)

Rashid, F.; Ghaffar, A.

2011-01-01

The cesium-137 is the most problematic radionuclide in the radioactive wastes. It belongs to the IA group of the periodic table, highly reactive towards water and has very high mobility. Due to beta and gamma radiation hazards of radiocesium its decontamination and disposal requires some special tools and techniques. In this study globules of clay material was used for the removal of cesium from low level liquid radioactive wastes and further processed for immobilization. The aim of this study was to assess the solidification and immobilization of secondary waste. The secondary waste, after sorption of cesium from the liquid radioactive waste generated at this institute, was found compatible to the cement matrix used for the cementation process. The procedure for immobilization of low level radioactive waste with cementation using vitreous clay material as an additive was developed. (orig.)

Radioactive liquid waste solidifying device

International Nuclear Information System (INIS)

Uchiyama, Yoshio.

1987-01-01

Purpose: To eliminate the requirement for discharge gas processing and avoid powder clogging in a facility suitable to the volume-reducing solidification of regenerated liquid wastes containing sodium sulfate. Constitution: Liquid wastes supplied to a liquid waste preheater are heated under a pressure higher than the atmospheric pressure at a level below the saturation temperature for that pressure. The heated liquid wastes are sprayed from a spray nozzle from the inside of an evaporator into the super-heated state and subjected to flash distillation. They are further heated to deposit and solidify the solidification components in the solidifying evaporation steams. The solidified powder is fallen downwardly and heated for removing water content. The recovered powder is vibrated so as not to be solidified and then reclaimed in a solidification storage vessel. Steams after flash distillation are separated into gas, liquid and solids by buffles. (Horiuchi, T.)

Radioactive waste management

International Nuclear Information System (INIS)

Kawakami, Yutaka

2008-01-01

Radioactive waste generated from utilization of radioisotopes and each step of the nuclear fuel cycle and decommissioning of nuclear facilities are presented. On the safe management of radioactive waste management, international safety standards are established such as ''The Principles of Radioactive Waste Management (IAEA)'' and T he Joint Convention on the Safety of Radioactive Waste Management . Basic steps of radioactive waste management consist of treatment, conditioning and disposal. Disposal is the final step of radioactive waste management and its safety is confirmed by safety assessment in the licensing process. Safety assessment means evaluation of radiation dose rate caused by radioactive materials contained in disposed radioactive waste. The results of the safety assessment are compared with dose limits. The key issues of radioactive waste disposal are establishment of long term national strategies and regulations for safe management of radioactive waste, siting of repository, continuity of management activities and financial bases for long term, and security of human resources. (Author)

Development of volume reduction treatment techniques for low level radioactive wastes

International Nuclear Information System (INIS)

Nabatame, Yasuzi

1984-01-01

The solid wastes packed in drums are preserved in the stores of nuclear establishments in Japan, and the quantity of preservation has already reached about 60 % of the capacity. It has become an important subject to reduce the quantity of generation of radioactive wastes and how to reduce the volume of generated wastes. As the result of the research aiming at the development of the solidified bodies which are excellent in the effect of volume reduction and physical properties, it was confirmed that the plastic solidified bodies using thermosetting resin were superior to conventional cement or asphalt solidification. The plastic solidifying system can treat various radioactive wastes. After radioactive wastes are dried and powdered, they are solidified with plastics, therefore, the effect of volume reduction is excellent. The specific gravity, strength and the resistance to water, fire and radiation were confirmed to be satisfacotory. The plastic solidifying system comprises three subsystems, that is, drying system, powder storing and supplying system and plastic solidifying system. Also the granulation technique after drying and powdering, acid decomposition technique, the microwave melting and solidifying technique for incineration ash, plasma melting process and electrolytic polishing decontamination are described. (Kako, I.)

Design and installation of a laboratory-scale system for radioactive waste treatment

International Nuclear Information System (INIS)

Berger, D.N.; Knox, C.A.; Siemens, D.H.

1980-05-01

Described are the mechanical design features and remote installation of a laboratory-scale radiochemical immobilization system which is to provide a means at Pacific Northwest Laboratory of studying effluents generated during solidification of high-level liquid radioactive waste. Detailed are the hot cell, instrumentation, two 4-in. and 12-in. service racks, the immobilization system modules - waste feed, spray calciner unit, and effluent - and a gamma emission monitor system for viewing calcine powder buildup in the spray calciner/in-can melter

The solidification of aluminum production waste in geopolymer matrix

Czech Academy of Sciences Publication Activity Database

Perná, Ivana; Hanzlíček, Tomáš

2014-01-01

Roč. 84, DEC 1 (2014), s. 657-662 ISSN 0959-6526 Institutional support: RVO:67985891 Keywords : aluminum waste * solidification * recycling * geopolymer Subject RIV: DM - Solid Waste and Recycling Impact factor: 3.844, year: 2014

The management of radioactive wastes in China

International Nuclear Information System (INIS)

Teng Lijun

2001-01-01

is used in Qinshan and Daya Bay NPPs. The bituminization is used for the LLLW treatment. Up to now, some LLLW has been solidified by using bituminization. The bituminized lumps (200L steel tanks) produced by solidification are now stored in repository. R and D work on waste packing containers has been carried in China, which include steel barrel, steel and concrete cases. A series of standards have been established for these containers. Disposal of radioactive wastes. A large mount of ILLW is stored in the underground carbon steel tanks after evaporation, concentration and neutralization. Through research and demonstration for more than ten years, it was decided that by using hydraulic fracturing and large volume pouring, the medium-level concentrated waste was disposed of in Sichuan and Lanzhou separately. According to the policy on LILW regional disposal, the national regional repositories have been constructed in the area where nuclear facilities are comparatively concentrated to dispose of local LILW. So far there are 2 repositories in China, the Northwest Repository and Guangdong Beilong Repository. During the 1980's, China has engaged in the research on the HLW disposal, the planning on the HLW deep geological disposal has been formulated. The multidisciplinary studies involving geology, hydrogeology, etc. have been carried in the Beishan area in Gansu Province. For application waste, there are some storage facilities all over country. (author)

Mathematical modelling of transport phenomena in radioactive waste-cement-bentonite matrix

International Nuclear Information System (INIS)

Plecas, Ilija; Dimovic, Slavko

2010-01-01

the volume of the waste (e.g. by incineration or compaction); solidifying non-solid wastes to make them physically stable, and packaging the waste to isolate it from the environment. Ion exchange resins may be used most successfully for the removal of radioactive and stable ions from dilute solutions. Ion exchange resins are polymers with cross-linking (connections between long carbon chains in a polymer). The resin has active groups in the form of electrically charged sites. At these sites ions of opposite charge are attracted but may be replaced by other ions depending on their relative concentrations and affinities for the sites. Spent mix bead exchange resins containing 60 Co and 137 Cs, represent a major portion of the solid radioactive waste in nuclear technology. Cement is used as a solidification material for the storage of intermediate-level radioactive waste. However, the retention of cesium, in the cement matrix is negligible. The sorption of cesium on cement is low and diffusivity of cesium in the hydrated cement is high. Only when the cement is mixed with a material having a significant sorption capacity, normally bead or powdered ion exchange resins, is the leachability of cesium and cobalt from the cement matrix low enough to be acceptable. Although cement has several unfavorable characteristics as a solidifying material, i.e. low volume reduction and relatively high leachability, it possesses many practical advantages: good mechanical characteristics, low cost, easy operation and radiation and thermal stability. It is generally assumed that the cement leachability of 137 Cs and other radionuclides can be reduced by adding minerals like bentonite, vermiculite and zeolite. At the 'Vinca' Institute of Nuclear Sciences, a promising composite for the solidification of radioactive wastes has been developed. Leaching of 137 Cs from this material was studied using the method recommended by the IAEA. (authors)

Feasibility study of solidification for low-level liquid waste generated by sulfuric acid elution treatment of spent ion exchange resin

International Nuclear Information System (INIS)

Asano, Takashi; Kawasaki, Tooru; Higuchi, Natsuko; Horikawa, Yoshihiko

2008-01-01

We studied cement-like solidification process for low-level liquid waste with relatively high levels of radioactivity that contains a high concentration of sodium sulfate. For this type waste, it is important that the sulfate ion should not dissolve from the solid waste because it forms ettringite on reaction with minerals in the concrete of the planned repository, and this leads to cracking during repository storage. It is also preferable that the pH of the pore water of the solid waste be low, because the bentonite of the repository changes in quality on exposure to alkaline solution. Therefore, the present solidification process has two procedures: conversion into insoluble sulfate from sodium sulfate (CIS) and formation of low pH cement-like solid (FLS). In the CIS procedure, BaSO 4 precipitation occurs with addition of Ba(OH) 2 ·8H 2 O to the liquid waste. In the FLS procedure, silica fume and blast furnace slag are added to the liquid waste containing BaSO 4 precipitate. We show the range of appropriate Ba/SO 4 molar ratio is from 1.1 to 1.5 in the present solidification process by leaching tests for some kinds of solid waste samples. The CIS reaction yield is over 98% at a typical CIS condition, i.e. Ba/SO 4 molar ratio=1.3, reaction temperature=60 deg C, and time=3 hr. (author)

The incineration of solid radioactive waste: a centralized solution

International Nuclear Information System (INIS)

Hernborg, G.; Broden, K.; Oehrn, G.

1985-01-01

Almost all the combustible low-level β- and γ-radioactive waste from Sweden, and even some waste from German nuclear power plants, is treated in an incineration plant at Studsvik. To date most of the ash has been put into 100-litre drums, which in turn have been put in 200-litre drums with concrete in between. Recently, methods have been developed and equipment installed for homogeneous solidification of the ash into concrete. Over the years since the start-up of the plant in 1976 the incinerator has worked with a high availability factor. Personnel doses and activity releases to the environment are well below limits set by regulatory authorities. (orig.)

Biodegradation of radioactive organic liquid waste from spent fuel reprocessing

International Nuclear Information System (INIS)

Ferreira, Rafael Vicente de Padua

2008-01-01

The research and development program in reprocessing of low burn-up spent fuel elements began in Brazil in 70's, originating the lab-scale hot cell, known as Celeste located at Nuclear and Energy Research Institute, IPEN - CNEN/SP. The program was ended at the beginning of 90's, and the laboratory was closed down. Part of the radioactive waste generated mainly from the analytical laboratories is stored waiting for treatment at the Waste Management Laboratory, and it is constituted by mixture of aqueous and organic phases. The most widely used technique for the treatment of radioactive liquid wastes is the solidification in cement matrix, due to the low processing costs and compatibility with a wide variety of wastes. However, organics are generally incompatible with cement, interfering with the hydration and setting processes, and requiring pre -treatment with special additives to stabilize or destroy them. The objective of this work can be divided in three parts: organic compounds characterization in the radioactive liquid waste; the occurrence of bacterial consortia from Pocos de Caldas uranium mine soil and Sao Sebastiao estuary sediments that are able to degrade organic compounds; and the development of a methodology to biodegrade organic compounds from the radioactive liquid waste aiming the cementation. From the characterization analysis, TBP and ethyl acetate were chosen to be degraded. The results showed that selected bacterial consortia were efficient for the organic liquid wastes degradation. At the end of the experiments the biodegradation level were 66% for ethyl acetate and 70% for the TBP. (author)

Method of processing liquid waste containing fission product

International Nuclear Information System (INIS)

Funabashi, Kiyomi; Kawamura, Fumio; Matsuda, Masami; Komori, Itaru; Miura, Eiichi.

1988-01-01

Purpose: To prepare solidification products of low surface dose by removing cesium which is main radioactive nuclides from re-processing plants. Method: Liquid wastes containing a great amount of fission products are generated accompanying the reprocessing for spent nuclear fuels. After pH adjustment, the liquid wastes are sent to a concentrator to concentrate the dissolved ingredients. The concentrated liquid wastes are pumped to an adsorption tower in which radioactive cesium contributing much to the surface dose is removed. Then, the liquid wastes are sent by way of a surge tank to a mixing tank, in which they are mixed under stirring with solidifying agents such as cements. Then, the mixture is filled in a drum-can and solidified. According to this invention, since radioactive cesium is removed before solidification, it is possible to prepare solidification products at low surface dose and facilitate the handling of the solidification products. (Horiuchi, T.)

Radioactive wastes from the nuclear fuel cycle

International Nuclear Information System (INIS)

Tomlinson, R.E.

1975-01-01

This symposium of 29 papers covers the following topics: overview; management and regulatory aspects; processing and solidification of low-level liquids; treatment of low-level solids; concentration and storage of high-level liquid wastes; and, solidification and storage of high level wastes. Selected papers are indexed separately

Preliminary tests of an infrared process monitor for polyethylene encapsulation of radioactive waste

International Nuclear Information System (INIS)

Wright, S.L.; Jones, R.W.; McClelland, J.F.; Kalb, P.D.

1996-01-01

Polyethylene encapsulation is a process that is being investigated for the solidification of radioactive nitrate salts at Brookhaven National Laboratory and Rocky Flats Plant. In the encapsulation process, radioactive-salt waste is mixed with polyethylene pellets, heated, and extruded as a molten stream. Upon cooling, the mixture solidifies to a monolithic waste form with excellent properties for long-term waste storage. This paper describes a novel method to monitor the composition of the salt/polymer stream as it exits the extruder. The monitor is based on a technique known as transient infrared spectroscopy (TIRS). The TIRS monitor is able to capture the real-time mid-infrared spectrum of the processed waste stream as it exits the extruder. The wealth of chemical information contained in a mid-infrared spectrum makes this technique very appealing for on-line monitoring and process control. Data from the monitor can be used to guide processing, minimize waste volume, and certify the composition of the final waste form

A comparison and cross-reference of commercial low-level radioactive waste acceptance criteria

International Nuclear Information System (INIS)

Kerr, T.A.

1997-04-01

This document, prepared by the National Low-Level Waste Management Program at the Idaho National Engineering and Environmental Laboratory, is a comparison and cross-reference of commercial low-level radioactive waste acceptance criteria. Many of these are draft or preliminary criteria as well as implemented criteria at operating low-level radioactive waste management facilities. Waste acceptance criteria from the following entities are included: US Nuclear Regulatory Commission, South Carolina, Washington, Utah, Nevada, California, illinois, Texas, North Carolina, Nebraska, Pennsylvania, New York, and the Midwest Compact Region. Criteria in the matrix include the following: physical form, chemical form, liquid limits, void space in packages, concentration averaging, types of packaging, chelating agents, solidification media, stability requirements, sorptive media, gas, oil, biological waste, pyrophorics, source material, special nuclear material, package dimensions, incinerator ash, dewatered resin, transuranics, and mixed waste. Each criterion in the matrix is cross-referenced to its source document so that exact requirements can be determined

Low-level radioactive waste, mixed low-level radioactive waste, and biomedical mixed waste

International Nuclear Information System (INIS)

Anon.

1994-01-01

This document describes the proceedings of a workshop entitled: Low-Level Radioactive Waste, Mixed Low-Level Radioactive Waste, and Biomedical Mixed Waste presented by the National Low-Level Waste Management Program at the University of Florida, October 17-19, 1994. The topics covered during the workshop include technical data and practical information regarding the generation, handling, storage and disposal of low-level radioactive and mixed wastes. A description of low-level radioactive waste activities in the United States and the regional compacts is presented

Radioactive wastes

International Nuclear Information System (INIS)

Teillac, J.

1988-01-01

This study of general interest is an evaluation of the safety of radioactive waste management and consequently the preservation of the environment for the protection of man against ionizing radiations. The following topics were developed: radiation effects on man; radioactive waste inventory; radioactive waste processing, disposal and storage; the present state and future prospects [fr

Contributions of the Nuclear Research Institute to the French-Czechoslovak seminar on the management of radioactive wastes held on 12-14 May, 1987

International Nuclear Information System (INIS)

1987-01-01

Paper were submitted on the use of calcination in liquid radioactive waste solidification; experience with the operation of mobile lines of the MESA type which are tested at nuclear power plants; the treatment of low level liquid wastes from special laundries. Other papers described experience with the operation of the facility for processing low and intermediate level wastes run by UJV (Nuclear Research Institute) since 1962, and the conditions for a radioactive waste burial site in Czechoslovakia. (E.S.). 3 tabs

Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes

International Nuclear Information System (INIS)

1979-01-01

Objective was to review the relative merits and potential of eleven alternative waste forms being considered for the solidification and disposal of radioactive wastes. A numerical rating of the alternative waste forms was arrived at individually by peer review panel members taking into consideration nine scientific and nine engineering parameters affecting the long-term performance and production of waste forms. A group rating for the alternative forms was achieved by averaging the individiual scores and discussing the available data base. Three final ranking lists comparing: (A) Present Scientific Merits or Least Risk for Use Today; (B) Research Priority; and (3) Present and Potential Engineering Practicality were prepared by the Panel. Each waste form in the lists is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or (3) Bottom Rank. Relative strengths and weaknesses of the alternative waste forms and recommendations for future program directions are discussed

Ecological solution of the problem of handling liquid radioactive wastes - Lr (by the example of Flue SSC RF RIA R)

International Nuclear Information System (INIS)

Polyakov, V.I.; Bukvich, B.A.

2006-01-01

A sharp reduction of nuclear waste amounts is possible if their elements are considered as source material of atomic complexes - SMAC. The prospect of their possible salvaging will require technological changes and ensuring safety of storage of the material till the need arises. Long experience in deep liquid radioactive waste disposal and accounting, calculations, and motivations demonstrate that a corresponding choice of geological formations makes it possible to abandon liquid radioactive waste solidification and ensure their isolation from environment when the most rigid radiation safety requirements are fulfilled. (author)

Sulfur polymer cement, a solidification and stabilization agent for hazardous and radioactive wastes

International Nuclear Information System (INIS)

Darnell, G.R.

1992-01-01

Hydraulic cements have been the primary radioactive waste stabilization agents in the United States for 50 years. Twelve years ago, Brookhaven National Laboratory was funded by the Department of Energy's Defense Low-Level Waste Management Program to test and develop sulfur polymer cement (SPC). It has stabilized routine wastes as well as some troublesome wastes with high waste-to-agent ratios. The Department of Energy's Hazardous Waste Remedial Action Program joined the effort by providing funding for testing and developing sulfur polymer cement as a hazardous-waste stabilization agent. Sulfur polymer cement has passed all the laboratory scale tests required by the US Environmental Protection Agency and US Nuclear Regulatory Commission. Two decades of tests by the US Bureau of Mines and private concrete contractors indicate this agent is likely to exceed other agents in longevity. This bulletin provides technical data from pertinent tests conducted by these various entities

Classification of solid wastes as non-radioactive wastes

International Nuclear Information System (INIS)

Suzuki, Masahiro; Tomioka, Hideo; Kamike, Kozo; Komatu, Junji

1995-01-01

The radioactive wastes generally include nuclear fuels, materials contaminated with radioactive contaminants or neutron activation to be discarded. The solid wastes arising from the radiation control area in nuclear facilities are used to treat and stored as radioactive solid wastes at the operation of nuclear facilities in Japan. However, these wastes include many non-radioactive wastes. Especially, a large amount of wastes is expected to generate at the decommissioning of nuclear facilities in the near future. It is important to classify these wastes into non-radioactive and radioactive wastes. The exemption or recycling criteria of radioactive solid wastes is under discussion and not decided yet in Japan. Under these circumstances, the Nuclear Safety Committee recently decided the concept on the category of non-radioactive waste for the wastes arising from decommissioning of nuclear facilities. The concept is based on the separation and removal of the radioactively contaminated parts from radioactive solid wastes. The residual parts of these solid wastes will be treated as non-radioactive waste if no significant difference in radioactivity between the similar natural materials and materials removed the radioactive contaminants. The paper describes the procedures of classification of solid wastes as non-radioactive wastes. (author)

Method of volume-reducing processing for radioactive wastes

International Nuclear Information System (INIS)

Sato, Koei; Yamauchi, Noriyuki; Hirayama, Toshihiko.

1985-01-01

Purpose: To process the processing products of radioactive liquid wastes and burnable solid wastes produced from nuclear facilities into stable solidification products by heat melting. Method: At first, glass fiber wastes of contaminated air filters are charged in a melting furnace. Then, waste products obtained through drying, sintering, incineration, etc. are mixed with a proper amount of glass fibers and charged into the melting furnace. Both of the charged components are heated to a temperature at which the glass fibers are melted. The burnable materials are burnt out to provide a highly volume-reduced products. When the products are further heated to a temperature at which metals or metal oxides of a higher melting point than the glass fiber, the glass fibers and the metals or metal oxides are fused to each other to be combined in a molecular structure into more stabilized products. The products are excellent in strength, stability, durability and leaching resistance at ambient temperature. (Kamimura, M.)

Method of producing solidification product of radioactive waste

International Nuclear Information System (INIS)

Masuda, Shunji; Iwami, Etsuji; Kadota, Keishi.

1989-01-01

Layers of thermosetting resin composition capable of curing at normal temperature are formed to a thickness of 2 to 5 mm at the bottom of a container. As the thermosetting resin composition capable of curing at normal temperature, there can be mentioned, for example, unsaturated polyester resin comprising a polymerizable monomer and an unsaturated polyester. After the layers are cured, a mixture of radioactive wastes and the thermosetting resin composition capable of curing at normal temperature is filled on the layer. After curing, thermosetting resin composition capable of curing at normal temperature is filled so as to fill gaps between the curing product and the container caused by curing shrinkage and at the upper surface of the curing products. After curing, plastic layers are formed at the surface. This can avoid residual bubbles in the layers or development of cracks. Further, leaching rate of Na ions is low and water proofness can be improved as well. (T.M.)

Development of a technology and a pilot plant for treatment of small volumes of liquid radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Stefanova, I G; Gradev, G D [Bulgarian Academy of Sciences, Sofia (Bulgaria). Inst. for Nuclear Research and Nuclear Energy

1997-02-01

The development of technology for treatment of liquid radioactive waste is described. Waste arisings are estimated. Liquid wastes of concern are mainly low active wastes according to the Bulgarian legislation. The activity is determined by the presence of {sup 134}Cs, {sup 137}Cs, {sup 60}Co, {sup 90}Sr, {sup 144}Ce, {sup 65}Zc, {sup 54}Mn, {sup 110m}Ag. Different precipitation processes are compared. The mixed iron hydroxide - calcium phosphate precipitation is determined as suitable for decontamination of the liquid radioactive waste. Effective decontamination is achieved when precipitation is followed by ion exchange. Additional increase of the decontamination is possible when sorbents are added during the precipitation step. The sorption and desorption of radionuclides on zeolites are studied. Cement solidification and thermal treatment of zeolites are studied for immobilization of radioactive material from precipitation and ion exchange. Both methods produce stable waste forms suitable for containment of the radionuclides. (author). 17 refs, 3 figs, 12 tabs.

Analysis of capital and operating costs associated with high level waste solidification processes

International Nuclear Information System (INIS)

Heckman, R.A.; Kniazewycz, B.G.

1978-03-01

An analysis was performed to evaluate the sensitivity of annual operating costs and capital costs of waste solidification processes to various parameters defined by the requirements of a proposed Federal waste repository. Five process methods and waste forms examined were: salt cake, spray calcine, fluidized bed calcine, borosilicate glass, and supercalcine multibarrier. Differential cost estimates of the annual operating and maintenance costs and the capital costs for the five HLW solidification alternates were developed

Solidification as low cost technology prior to land filling of industrial hazardous waste sludge.

Science.gov (United States)

El-Sebaie, O; Ahmed, M; Ramadan, M

2000-01-01

The aim of this study is to stabilize and solidify two different treated industrial hazardous waste sludges, which were selected from factories situated close to Alexandria. They were selected to ensure their safe transportation and landfill disposal by reducing their potential leaching of hazardous elements, which represent significant threat to the environment, especially the quality of underground water. The selected waste sludges have been characterized. Ordinary Portland Cement (OPC), Cement Kiln Dust (CKD) from Alexandria Portland Cement Company, and Calcium Sulphate as a by-product from the dye industry were used as potential solidification additives to treat the selected treated waste sludges from tanning and dyes industry. Waste sludges as well as the solidified wastes have been leach-tested, using the General Acid Neutralization Capacity (GANC) procedure. Concentration of concerning metals in the leachates was determined to assess changes in the mobility of major contaminants. The treated tannery waste sludge has an acid neutralization capacity much higher than that of the treated dyes waste sludge. Experiment results demonstrated the industrial waste sludge solidification mix designs, and presented the reduction of contaminant leaching from two types of waste sludges. The main advantages of solidification are that it is simple and low cost processing which includes readily available low cost solidification additives that will convert industrial hazardous waste sludges into inert materials.

A analysis of cementation technology for liquid radioactive-waste in PWR NPPs

International Nuclear Information System (INIS)

Chen Liang; Chen Li; Li Junhua

2009-01-01

Cementation is one of the most popular solidification technology for the low-and-intermediate level liquid radioactive waste. It has been applied in all of domestic PWR NPPs. The process characteristics and operation of the cementations in the different NPPs are introduced,and the advantage and disadvantage of the cementation are analyzed in this paper. A drum and a cask are compared as a package of the solidified waste, the drum can decrease over 50% final volume of the waste, furthermore the cost for manufacture and transportation for this drum is more cheaper than the cask, but an additional shielding may be necessary for the waste with higher level radioactivity that is packed in drum. More waste can be contained if an appropriate in-drum mixer is used while secondary waste will be unavoidable if the out-drum mixing is adopted. A carriage can make it easier to decontaminate on the surface of equipment and on the floor, furthermore the carriage is more economical than a roller conveyor in manufacture and maintenance. The cementation recipe for the waste should be optimized and additive material should be as less as possible to increase the containing rate of the waste. (authors)

Species redistribution during solidification of nuclear fuel waste metal castings

Energy Technology Data Exchange (ETDEWEB)

Naterer, G F; Schneider, G E [Waterloo Univ., ON (Canada)

1994-12-31

An enthalpy-based finite element model and a binary system species redistribution model are developed and applied to problems associated with solidification of nuclear fuel waste metal castings. Minimal casting defects such as inhomogeneous solute segregation and cracks are required to prevent container corrosion and radionuclide release. The control-volume-based model accounts for equilibrium solidification for low cooling rates and negligible solid state diffusion for high cooling rates as well as intermediate conditions. Test problems involving nuclear fuel waste castings are investigated and correct limiting cases of species redistribution are observed. (author). 11 refs., 1 tab., 13 figs.

Solidification of liquid concentrate and solid waste generated as by-products of the liquid radwaste treatment systems in light-water reactors

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1977-01-01

The treatment of liquid concentrate and solid waste produced in light-water reactors as by-products of liquid radwaste treatment systems consists of five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging (solidification) and waste package handling. This paper will concern itself primarily with the solidification operation, however, the other operations enumerated as well as the types of wastes treated and their origins will be briefly described, especially with regards to their effects on solidification. During solidification, liquid concentrate and solid wastes are incorporated with a solidification agent to form a monolithic, free-standing solid. The basic solidification agent types either currently used in the United States or proposed for use include absorbants, hydraulic cement, urea-formaldehyde, other polymer systems, and bitumen. The operation, formulations and limitations of these agents as used for radwaste solidification will be discussed. Properties relevant to the evaluation of solidified waste forms will be identified and relative comparisons made for wastes solidified by various processes

The role of wastes from nuclear power plants in the overall management of radioactive wastes

International Nuclear Information System (INIS)

Krause, H.

1983-01-01

The wastes arising from nuclear power plants (NPP) are rather low in activity and the radionuclides contained therein have a low radiotoxicity and short half-life as a rule. However, NPPs are the largest in number among all nuclear facilities and produce the greatest amount of radioactive wastes. All NPPs have been able to keep the radiation doses in the environment below the permissible values, in most cases at 1 mrem/a or even lower. The methods applied for the treatment of liquid radioactive effluents have reached a high degree of effectivity and reliability. For the solidification of the residues several appropriate methods are available. However, some improvements are still desirable. Although methods exist for incineration, cutting and baling of solid wastes, only the last method is employed at NPPs as a rule. Central treatment facilities could improve this situation. The exhaust air treatment has reached a state that satisfies high standards during normal operation and in design base accidents. Improvements seem indicated regarding the in situ-inspection of HEPA-filters and the protection of filters against excess humidity and droplets. The partial and total decommissioning of NPPs has already been demonstrated. The problems caused by the wastes arising from such actions are in the same range as those from routine operation and maintenance of NPPs. Large amounts of radioactive wastes have already been disposed of by shallow land burial, disposal into deep geological formations or dumping into the deep sea. Specific standards could probably facilitate the disposal of wastes from NPPs. The present management of radioactive wastes from NPPs satisfies all actual needs. Therefore, spectacular new developments are neither required nor to be expected. However, by the continuous improvement of details and by optimization of the whole system progress can still be achieved and useful contributions to the further development of nuclear energy be made. (author)

Processing method for cleaning water waste from cement kneader

International Nuclear Information System (INIS)

Soda, Kenzo; Fujita, Hisao; Nakajima, Tadashi.

1990-01-01

The present invention concerns a method of processing cleaning water wastes from a cement kneader in a case of processing liquid wastes containing radioactive wastes or deleterious materials such as heavy metals by means of cement solidification. Cleaning waste wastes from the kneader are sent to a cleaning water waste tank, in which gentle stirring is applied near the bottom and sludges are retained so as not to be coagulated. Sludges retained at the bottom of the cleaning water waste tank are sent after elapse of a predetermined time and then kneaded with cements. Thus, since the sludges in the cleaning water are solidified with cement, inhomogenous solidification products consisting only of cleaning sludges with low strength are not formed. The resultant solidification product is homogenous and the compression strength thereof reaches such a level as capable of satisfying marine disposal standards required for the solidification products of radioactive wastes. (I.N.)

Radioactive waste management solutions

International Nuclear Information System (INIS)

Siemann, Michael

2015-01-01

One of the more frequent questions that arise when discussing nuclear energy's potential contribution to mitigating climate change concerns that of how to manage radioactive waste. Radioactive waste is produced through nuclear power generation, but also - although to a significantly lesser extent - in a variety of other sectors including medicine, agriculture, research, industry and education. The amount, type and physical form of radioactive waste varies considerably. Some forms of radioactive waste, for example, need only be stored for a relatively short period while their radioactivity naturally decays to safe levels. Others remain radioactive for hundreds or even hundreds of thousands of years. Public concerns surrounding radioactive waste are largely related to long-lived high-level radioactive waste. Countries around the world with existing nuclear programmes are developing longer-term plans for final disposal of such waste, with an international consensus developing that the geological disposal of high-level waste (HLW) is the most technically feasible and safe solution. This article provides a brief overview of the different forms of radioactive waste, examines storage and disposal solutions, and briefly explores fuel recycling and stakeholder involvement in radioactive waste management decision making

Radioactive waste management at the Paul Scherrer Institute, the largest Swiss national research centre

Directory of Open Access Journals (Sweden)

Beer Hans-Frieder

2009-01-01

Full Text Available This paper presents the current radioactive waste management practices at the Paul Scherrer Institute (PSI. The PSI contributes to waste related problems in two aspects, namely to the scientific basis of waste management and disposal, and to the practical treatment and storage of radioactive waste. In addition to the tasks of treating on-site generated waste, PSI manages the wastes from medicine, industry, and research throughout Switzerland on behalf of the government. Therefore the Dismantling and Waste Management Section is a part of the Logistics Department at PSI. Proved and accepted methods have to be developed for the safe conditioning and storage of radioactive waste. Various waste treatment facilities exist at PSI. The conditioning facility is dedicated to sorting, compaction by a 120 t press, solidification with special cement, and embedding in concrete. Specialized facilities were constructed for waste from the decommissioning of research reactors. Activated aluminum and its alloys were melted in crucibles and embedded in concrete in a concrete container. After dismantling the structural material of the reactors, it was embedded in concrete in the same manner. For the conditioning of activated reactor graphite, a dedicated method was developed. Graphite was crushed to replace sand in the grout, for embedding radioactive waste in concrete containers. For accelerator waste, a walk-in hot cell equipped with an electrically driven manipulator is available where the highly activated large components (targets, beam dump can be cut into pieces and embedded in concrete in containers. To guarantee the fulfillment of the demands of the regulators, the Dismantling and Waste Management Section applies an accredited quality management system for the safe collection, conditioning, and storage of radioactive waste.

The system for centralized inventory keeping and ultimate disposal of radioactive waste in the former German Democratic Republic

International Nuclear Information System (INIS)

Beise, E.; Mielke, H.G.; Mueller, W.; Oppermann, U.

1991-01-01

The report explains the concept adopted by the former GDR. The system based at Morsleben, for centralized inventory keeping and management of radioactive waste is explained, refewing to the amounts of waste accrued, storage and transport of waste drums, classification and preparation of waste forms, and ultimate disposal of radioactive waste in the Morsleben repository. The report includes information on the management of special waste and spent fuel elements which cannot be stored at the Morsleben site. Most of the radioactive waste produced in the former GDR has been stored since 1979 at the Morsleben site. The waste came from the nuclear power plants (Greifswald, Rheinsberg), and from installations and institutes applying or producing radionuclides - so-called APR waste - (e.g. from the institutes at Rossendorf and Berlin-Buch, and from about 1300 other waste producers). The waste was accepted as or processed to solid waste forms, liquid waste, sealed radiation sources, and special waste; the ultimate storage techniques applied are packing of drums, backfilling, solidification of liquid waste and disposal in boreholes. Up to the end of the year 1989, the Morsleben repository received about 14000 m 3 of radioactive waste (about 40% solid waste, and about 60% liquid waste). (orig.) [de

Disposal of radioactive wastes

International Nuclear Information System (INIS)

Blomeke, J.O.

1979-01-01

Radioactive waste management and disposal requirements options available are discussed. The possibility of beneficial utilization of radioactive wastes is covered. Methods of interim storage of transuranium wastes are listed. Methods of shipment of low-level and high-level radioactive wastes are presented. Various methods of radioactive waste disposal are discussed

Radioactive wastes

International Nuclear Information System (INIS)

Grass, F.

1982-01-01

Following a definition of the term 'radioactive waste', including a discussion of possible criteria allowing a delimitation of low-level radioactive against inactive wastes, present techniques of handling high-level, intermediate-level and low-level wastes are described. The factors relevant for the establishment of definitive disposals for high-level wastes are discussed in some detail. Finally, the waste management organization currently operative in Austria is described. (G.G.)

Radioactive waste management

International Nuclear Information System (INIS)

Balek, V.

1994-01-01

This booklet is a publication by International Atomic Energy Agency for general awareness of citizens and policy-makers to clarify their concept of nuclear wastes. In a very simple way it tells what is radioactivity, radiations and radioactive wastes. It further hints on various medial and industrial uses of radiations. It discusses about different types of radioactive wastes and radioactive waste management. Status of nuclear power plants in Central and Eastern European countries are also discussed

Defense waste solidification studies, 200-S area. Savannah River Plant work request 860504, Project S-1780

International Nuclear Information System (INIS)

1977-05-01

A scope of work and a venture guidance appraisal were prepared for a conceptual process and plant facilities for the solidification and long-term storage of radioactive wastes removed from underground storage tanks in the 241 F and H Areas at the Savannah River Plant. Conceptual design was based on incorporating the highly radioactive waste components in a borosilicate type glass. The scope of work describes facilities for: reclaiming liquid and sludge wastes from F and H area tank farms; separating the sludge from the liquid salt solution by physical processes; removing radioactive cesium from the salt solution by ion exchange techniques; incorporating the dried sludge and cesium in a borosilicate glass in stainless steel containers; evaporating the liquid salt solution and encapsulating the resulting salt cake in a stainless steel container; and storing two years' worth of glass and salt containing cyclinders in separate retrievable surface storage facilities. Operations are to be located in a new area, designated the 200-S area. A full complement of power, general, and service facilities are provided. The venture guidance appraisal based on FY 82 authorization and FY 87 turnover is $2,900,000,000. The figure is suitable for planning purposes only. The Glass-form Waste Case is a variation of the concrete-form waste case (or the Reference Plant Case) reported in DPE--3410. The new venture guidance appraisal for the concrete-form case (updated to a consistent time basis with the glass-form case) is $2,900,000,000, indicating no apparent cost advantage between the two waste product forms

Understanding radioactive waste

International Nuclear Information System (INIS)

Murray, R.L.

1981-12-01

This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes)

Understanding radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Murray, R.L.

1981-12-01

This document contains information on all aspects of radioactive wastes. Facts are presented about radioactive wastes simply, clearly and in an unbiased manner which makes the information readily accessible to the interested public. The contents are as follows: questions and concerns about wastes; atoms and chemistry; radioactivity; kinds of radiation; biological effects of radiation; radiation standards and protection; fission and fission products; the Manhattan Project; defense and development; uses of isotopes and radiation; classification of wastes; spent fuels from nuclear reactors; storage of spent fuel; reprocessing, recycling, and resources; uranium mill tailings; low-level wastes; transportation; methods of handling high-level nuclear wastes; project salt vault; multiple barrier approach; research on waste isolation; legal requiremnts; the national waste management program; societal aspects of radioactive wastes; perspectives; glossary; appendix A (scientific American articles); appendix B (reference material on wastes). (ATT)

Overpack for processing radioactive waste

International Nuclear Information System (INIS)

Asano, Hidekazu.

1997-01-01

A glass solidification material in which radioactive wastes are sealed and solidified in glass is covered by an inner layer vessel made of corrosion resistant materials, and the outer side thereof is covered with an outer layer vessel made of a reinforced material. The inner layer vessel made of corrosion materials comprises corrosion materials such as titanium, copper, stainless steel and nickel based alloy, and the outer layer vessel made of a reinforced material comprises a reinforced material such as carbon steel. If it is constituted by using carbon steel having a thickness as much as of from 50 to 200mm, it is durable sufficiently under ground of about 1000m. Although the outer layer vessel made of the reinforced material is corroded by oxidation by oxygen contained in underwater after lapse of time of several years, it is endurable sufficiently to initial oxidative corrosion by determined the thickness to 50mm or more, and after oxygen is consumed, reductive corrosion with extremely slow progressing speed begins. Since the inner vessel made of the corrosion resistant material is formed, the lifetime is extended, and the glass solidification materials can be confined stably for a long period of time. (N.H.)

Radioactive wastes. Management

International Nuclear Information System (INIS)

Guillaumont, R.

2001-01-01

Many documents (journal articles, book chapters, non-conventional documents..) deal with radioactive wastes but very often this topic is covered in a partial way and sometimes the data presented are contradictory. The aim of this article is to precise the definition of radioactive wastes and the proper terms to describe this topic. It describes the main guidelines of the management of radioactive wastes, in particular in France, and presents the problems raised by this activity: 1 - goal and stakes of the management; 2 - definition of a radioactive waste; 3 - radionuclides encountered; 4 - radio-toxicity and radiation risks; 5 - French actors of waste production and management; 6 - French classification and management principles; 7 - wastes origin and characteristics; 8 - status of radioactive wastes in France per categories; 9 - management practices; 10 - packages conditioning and fabrication; 11 - storage of wastes; 12 - the French law from December 30, 1991 and the opportunities of new ways of management; 13 - international situation. (J.S.)

A literature review of mixed waste components: Sensitivities and effects upon solidification/stabilization in cement-based matrices

International Nuclear Information System (INIS)

Mattus, C.H.; Gilliam, T.M.

1994-03-01

The US DOE Oak Ridge Field Office has signed a Federal Facility Compliance Agreement (FFCA) regarding Oak Ridge Reservation (ORR) mixed wastes subject to the land disposal restriction (LDR) provisions of the Resource conservation and Recovery Act. The LDR FFCA establishes an aggressive schedule for conducting treatability studies and developing treatment methods for those ORR mixed (radioactive and hazardous) wastes listed in Appendix B to the Agreement. A development, demonstration, testing, and evaluation program has been initiated to provide those efforts necessary to identify treatment methods for all of the wastes that meet Appendix B criteria. The program has assembled project teams to address treatment development needs in a variety of areas, including that of final waste forms (i.e., stabilization/solidification processes). A literature research has been performed, with the objective of determining waste characterization needs to support cement-based waste-form development. The goal was to determine which waste species are problematic in terms of consistent production of an acceptable cement-based waste form and at what concentrations these species become intolerable. The report discusses the following: hydration mechanisms of Portland cement; mechanisms of retardation and acceleration of cement set-factors affecting the durability of waste forms; regulatory limits as they apply to mixed wastes; review of inorganic species that interfere with the development of cement-based waste forms; review of radioactive species that can be immobilized in cement-based waste forms; and review of organic species that may interfere with various waste-form properties

Topic of nuclear power plant wastes at fifth CMEA symposium on research of fuel and radioactive solution reprocessing

International Nuclear Information System (INIS)

Kyrs, M.; Napravnik, J.

1982-01-01

A survey is presented of the results of the work of Session 3 of the Symposium held in Marianske Lazne from April 7 to 10, 1981. The participants heard 44 papers mainly related to the following problem areas: existing methods and methods being developed of handling radioactive wastes from nuclear power plants; improvements in the technology of the concentration of liquid wastes by evaporation and other methods; solidification of concentrated liquid wastes into suitable form; methods of the treatment of solid (combustible, non-combustible and compactable) wastes; improvements in methods of the treatment of gaseous effluents. A survey was organized on criteria applied to methods used for radioactive waste processing. The inquiry showed that the principal criteria are the product Quality, costs for waste processing and the release of harmful substances into the environment. (Ha)

Low-level Radioactive waste Management

International Nuclear Information System (INIS)

1991-01-01

This meeting describes low-level radioactive waste management problems and contains 8 papers: 1 Low-level radioactive waste management: exemption concept and criteria used by international organizations. 2 Low-level radioactive waste management: french and foreign regulations 3 Low-level radioactive waste management in EDF nuclear power plants (FRANCE) 4 Low-level radioactive waste management in COGEMA (FRANCE) 5 Importance of low-level radioactive wastes in dismantling strategy in CEA (FRANCE) 6 Low-level radioactive waste management in hospitals 7 Low-level radioactive waste disposal: radiation protection laws 8 Methods of low-level radioactive materials measurements during reactor dismantling or nuclear facilities demolition (FRANCE)

Method of processing liquid wastes

International Nuclear Information System (INIS)

Naba, Katsumi; Oohashi, Takeshi; Kawakatsu, Ryu; Kuribayashi, Kotaro.

1980-01-01

Purpose: To process radioactive liquid wastes with safety by distillating radioactive liquid wastes while passing gases, properly treating the distillation fractions, adding combustible and liquid synthetic resin material to the distillation residues, polymerizing to solidify and then burning them. Method: Radioactive substance - containing liquid wastes are distillated while passing gases and the distillation fractions containing no substantial radioactive substances are treated in an adequate method. Synthetic resin material, which may be a mixture of polymer and monomer, is added together with a catalyst to the distillation residues containing almost of the radioactive substances to polymerize and solidify. Water or solvent in such an extent as not hindering the solidification may be allowed if remained. The solidification products are burnt for facilitating the treatment of the radioactive substances. The resin material can be selected suitably, methacrylate syrup (mainly solution of polymethylmethacrylate and methylmethacrylate) being preferred. (Seki, T.)

Radioactive waste processing

International Nuclear Information System (INIS)

Dejonghe, P.

1978-01-01

This article gives an outline of the present situation, from a Belgian standpoint, in the field of the radioactive wastes processing. It estimates the annual quantity of various radioactive waste produced per 1000 MW(e) PWR installed from the ore mining till reprocessing of irradiated fuels. The methods of treatment concentration, fixation, final storable forms for liquid and solid waste of low activity and for high level activity waste. The storage of radioactive waste and the plutonium-bearing waste treatement are also considered. The estimated quantity of wastes produced for 5450 MW(e) in Belgium and their destination are presented. (A.F.)

Remediation and decommissioning of radioactive waste facilities in Estonia

International Nuclear Information System (INIS)

Putnik, H.; Realo, E.

2001-01-01

Full text: The nuclear training facility at Paldiski was constructed in the early 1960's by the former USSR Navy. The hull sections of Delta and Echo class submarines each housing a full-sized ship reactor were installed in the main building of the site for training of navy personnel in safe operation of the submarine nuclear reactor systems. The first reactor was commissioned in 1968 and the second in 1982, while both was shut down in 1989. After Estonia's reproclamation of independence in 1991 the responsibility for the clean up and decommissioning of the Paldiski site became a subject of negotiations between Russia and Estonia. As the result Estonia took the ownership and control of the site in September 1995. Before the take over the Russian authorities defuelled the reactors and transported the spent fuel to Russia, dismantled the hull sections not related with reactor systems, seal-welded the hull sections housing the reactor vessels with their primary circuitry and enclosed those in reinforced concrete sarcophagi. The auxiliary facilities and radioactive waste were left intact. Main goals of the Conceptual Decommissioning Plan for the Paldiski facilities, developed under the auspices of the Paldiski International Expert Reference Group (Pier, a group established at the request of the Estonian government to advise local authorities to maintain the decommissioning and waste management at Paldiski) were defined as following: Establishing the waste management system and a long term monitored interim storage, corresponding to internationally accepted safety standards and capable to condition, receive and store all the waste generated during decommissioning of the facility; Reductions of the extent of radiologically controlled areas as much as possible, in order to minimise maintenance requirements. To achieve these goals the following main tasks were addressed in the short and medium term site management action plans: Rearrangement of site for the needs of

Radioactive wastes of Nuclear Industry

International Nuclear Information System (INIS)

1995-01-01

This conference studies the radioactive waste of nuclear industry. Nine articles and presentations are exposed here; the action of the direction of nuclear installations safety, the improvement of industrial proceedings to reduce the waste volume, the packaging of radioactive waste, the safety of radioactive waste disposal and environmental impact studies, a presentation of waste coming from nuclear power plants, the new waste management policy, the international panorama of radioactive waste management, the international transport of radioactive waste, finally an economic analysis of the treatment and ultimate storage of radioactive waste. (N.C.)

Radioactive Waste in Perspective

International Nuclear Information System (INIS)

2011-01-01

Large volumes of hazardous wastes are produced each year, however only a small proportion of them are radioactive. While disposal options for hazardous wastes are generally well established, some types of hazardous waste face issues similar to those for radioactive waste and also require long-term disposal arrangements. The objective of this NEA study is to put the management of radioactive waste into perspective, firstly by contrasting features of radioactive and hazardous wastes, together with their management policies and strategies, and secondly by examining the specific case of the wastes resulting from carbon capture and storage of fossil fuels. The study seeks to give policy makers and interested stakeholders a broad overview of the similarities and differences between radioactive and hazardous wastes and their management strategies. Contents: - Foreword; - Key Points for Policy Makers; - Executive Summary; - Introduction; - Theme 1 - Radioactive and Hazardous Wastes in Perspective; - Theme 2 - The Outlook for Wastes Arising from Coal and from Nuclear Power Generation; - Risk, Perceived Risk and Public Attitudes; - Concluding Discussion and Lessons Learnt; - Strategic Issues for Radioactive Waste; - Strategic Issues for Hazardous Waste; - Case Studies - The Management of Coal Ash, CO 2 and Mercury as Wastes; - Risk and Perceived Risk; - List of Participants; - List of Abbreviations. (authors)

Nuclear fuel cycle and waste management in France

International Nuclear Information System (INIS)

Sousselier, Yves.

1981-05-01

After a short description of the nuclear fuel cycle mining, milling, enrichment and reprocessing, radioactive waste management in France is exposed. The different types of radioactive wastes are examined. Storage, solidification and safe disposal of these wastes are described

Management of radioactive waste

International Nuclear Information System (INIS)

Neerdael, B.; Marivoet, J.; Put, M.; Van Iseghem, P.; Volckaert, G.; Wacquier, W.

1998-09-01

The document gives an overview of of different aspects of radioactive waste management in Belgium. The document discusses the radioactive waste inventory in Belgium, the treatment and conditioning of radioactive waste as well as activities related to the characterisation of different waste forms. A separate chapter is dedicated to research and development regarding deep geological disposal of radioactive waste. In the Belgian waste management programme, particular emphasis is on studies for disposal in clay. Main results of these studies are highlighted and discussed

Cement solidification of spent ion exchange resins produced by the nuclear industry

International Nuclear Information System (INIS)

Jaouen, C.; Vigreux, B.

1988-01-01

Cement solidification technology has been applied to spent ion exchange resins for many years in countries throughout the world (at reactors, research centers and spent fuel reprocessing plants). Changing specifications for storage of radioactive waste have, however, confronted the operators of such facilities with a number of problems. Problems related both to the cement solidification process (water/cement/resin interactions and chemical interactions) and to its utilization (mixing, process control, variable feed composition, etc.) have often led waste producers to prefer other, polymer-based processes, which are very expensive and virtually incompatible with water. This paper discusses research on cement solidification of ion exchange resins since 1983 and the development of application technologies adapted to nuclear service conditions and stringent finished product quality requirements

Immersed radioactive wastes

International Nuclear Information System (INIS)

2017-03-01

This document presents a brief overview of immersed radioactive wastes worldwide: historical aspects, geographical localization, type of wastes (liquid, solid), radiological activity of immersed radioactive wastes in the NE Atlantic Ocean, immersion sites and monitoring

Characterization on incineration residue of radioactive solid wastes

International Nuclear Information System (INIS)

Katoh, Kiyoshi; Hirayama, Katsuyoshi; Kato, Akira.

1989-01-01

Characterization was carried out on incineration residue discharged from the radioactive solid waste incineration unit (capacity, 100 kg/h) in use at the Tokai Research Establishment of Japan Atomic Energy Research Institute (JAERI) to obtain basic data for investigating solidification methods of the residue. The characterized residue was taken from furnace and a primary ceramic filter of the incineration unit which incinerates combustible solid wastes generated at JAERI and the outside organizations. Items of characterization involve a particle size distribution, misplaced materials content, ignition loss, chemical composition and radioactivity of nuclides in the ash. As the results, the size of ash sampled from the furnace distributed a wide range, with about 35∼60 % of ash smaller than 5 mm and about 10∼25 % of massive one larger than 30 mm (max. size: ∼130 mm). The ignition loss was 2∼3 %. The chemical compositions of the ash were mainly SiO 2 , Fe 2 O 3 , CaO and Al 2 O 3 . The specific activities of the ash were about 0.4∼4 x 10 3 Bq/g, and principal contaminants were 60 Co and 137 Cs. (author)

Management of Radioactive Wastes

International Nuclear Information System (INIS)

Tchokosa, P.

2010-01-01

Management of Radioactive Wastes is to protect workers and the public from the radiological risk associated with radioactive waste for the present and future. It application of the principles to the management of waste generated in a radioisotope uses in the industry. Any material that contains or is contaminated with radionuclides at concentrations or radioactivity levels greater than ‘exempt quantities’ established by the competent regulatory authorities and for which no further use is foreseen or intended. Origin of the Radioactive Waste includes Uranium and Thorium mining and milling, nuclear fuel cycle operations, Operation of Nuclear power station, Decontamination and decommissioning of nuclear facilities and Institutional uses of isotopes. There are types of radioactive waste: Low-level Waste (LLW) and High-level Waste. The Management Options for Radioactive Waste Depends on Form, Activity, Concentration and half-lives of the radioactive waste, Storage and disposal methods will vary according to the following; the radionuclides present, and their concentration, and radio toxicity. The contamination results basically from: Contact between radioactive materials and any surface especially during handling. And it may occur in the solid, liquid or gas state. Decontamination is any process that will either reduce or completely remove the amount of radionuclides from a contaminated surface

Radioactive waste management

International Nuclear Information System (INIS)

1992-01-01

This book highlights the main issues of public concern related to radioactive waste management and puts them into perspective. It provides an overview of radioactive waste management covering, among other themes, policies, implementation and public communication based on national experiences. Its purpose is to assists in increasing the understanding of radioactive waste management issues by public and national authorities, organizations involved in radioactive waste management and the nuclear industry; it may also serve as a source book for those who communicate with the public. Even in the unlikely event that nuclear power does not further develop around the world, the necessity for dealing with nuclear waste from past usages, from uranium mining and milling, decontamination and decommissioning of existing nuclear facilities and from the uses of radioactive materials in medicine, industry and research would still exist. In many countries, radioactive waste management planning involves making effective institutional arrangements in which responsibilities and liabilities are well established for the technical operation and long term surveillance of disposal systems. Financing mechanisms are part of the arrangements. Continuous quality assurance and quality control, at all levels of radioactive waste management, are essential to ensure the required integrity of the system. As with any other human activity, improvements in technology and economics may be possible and secondary problems avoided. Improvements and confirmation of the efficiency of processes and reduction of uncertainties can only be achieved by continued active research, development and demonstration, which are the goals of many national programmes. International co-operation, also in the form of reviews, can contribute to increasing confidence in the ongoing work. The problem of radioactive wastes is not a unique one; it may be compared with other problems of toxic wastes resulting from many other

Sulfur polymer cement, a new stabilization agent for mixed and low- level radioactive waste

International Nuclear Information System (INIS)

Darnell, G.R.

1991-01-01

Solidification and stabilization agents for radioactive, hazardous, and mixed wastes are failing to pass governmental tests at alarming rates. The Department of Energy's National Low-Level Waste Management Program funded testing of Sulfur Polymer Cement (SPC) by Brookhaven National Laboratory during the 1980s. Those tests and tests by the US Bureau of Mines (the original developer of SPC), universities, states, and the concrete industry have shown SPC to be superior to hydraulic cements in most cases. Superior in what wastes can be successfully combined and in the quantity of waste that can be combined and still pass the tests established by the US Environmental Protection Agency and the US Nuclear Regulatory Commission

Dynamics of radioactive waste generation

International Nuclear Information System (INIS)

Dogaru, Daniela; Virtopeanu, Cornelia; Ivan, Alexandrina

2008-01-01

In Romania there are in operation three facilities licensed for collection, treatment and storage of radioactive waste resulted from industry, research, medicine, and agriculture, named institutional radioactive waste. The repository, which is of near surface type, is designed for disposing institutional radioactive waste. The institutional radioactive wastes generated are allowed to be disposed into repository according to the waste acceptance criteria, defined for the disposal facility. The radioactive wastes which are not allowed for disposal are stored on the site of each facility which is special authorised for this. The paper describes the dynamics of generation of institutional waste in Romania, both for radioactive waste which are allowed to be disposed into repository and for radioactive waste which are not allowed to be disposed of. (authors)

Controlling radioactive waste

International Nuclear Information System (INIS)

Wurtinger, W.

1992-01-01

The guideline of the Ministry for Environmental Protection for controlling radioactive waste with a negligible development of heat defines in detail what data are relevant to the control of radioactive waste and should be followed up on and included in a system of documentation. By introducing the AVK (product control system for tracing the course of waste disposal) the operators of German nuclear power plants have taken the requirements of this guideline into account. In particular, possibilities for determining the degree of radioactivity of radioactive waste, which the BMU-guidelines call for, were put into practice by means of the programming technology of the product control system's module MOPRO. (orig.) [de

Aspects of radioactive waste management

International Nuclear Information System (INIS)

Cutoiu, Dan

2003-01-01

The origin and types of radioactive waste, the objective and the fundamental principles of radioactive waste management and the classification of radioactive waste are presented. Problems of the radioactive waste management are analyzed. (authors)

Radioactive waste management in nuclear power plants with WWER-type reactors

Energy Technology Data Exchange (ETDEWEB)

Dlouhy, Z; Napravnik, J; Safar, O

1975-05-01

The possibilities of radioactive waste solidification in nuclear power plants with LWR reactors (of the WWER type) and the problems of their safe storage in Czechoslovakia are discussed. The most suitable method for the treatment of emitted sorbents and concentrates seems to be their incorporation in bitumen or concrete. In the disposal of solidified blocks all requirements should be met including the selection of suitable sites and of convenient methods of transportation. A preliminary economic estimate shows that the storage of bitumen-incorporated wastes in trenches seems to be less expensive from the point of view of exploitation of the storage facility as well as from the point of view of investment.

Transport of radioactive wastes

International Nuclear Information System (INIS)

Stuller, C.

2003-01-01

In this article author describes the system of transport and processing of radioactive wastes from nuclear power of Slovenske elektrarne, plc. It is realized the assurance of transport of liquid and solid radioactive wastes to processing links from places of their formation, or of preliminary storage and consistent transports of treated radioactive wastes fixed in cement matrix of fibre-concrete container into Rebublic storage of radioactive wastes in Mochovce

Immobilization of radioactive and hazardous wastes in a newly developed sulfur polymer cement (Spc) matrix

International Nuclear Information System (INIS)

Abdel Raouf, M.W.; Husain, A.I.; El-Gammal, B.

2005-01-01

Low and Intermediate level radioactive wastes (LILW) and hazardous wastes, presents a waste disposal problem. In this respect, a process to immobilize different radioactive and hazardous wastes, including metals contaminated with radionuclides in a form that is non-dispersible and meet the Environmental Protection Agency (USA, EPA) leaching criteria is a must. In this stabilization and solidification process (S/S), simulated radioactive wastes of Cs, Sr, Ce, Cr, and Pb were immobilized in the stable form of sulfur polymer cement (SPC). In the present work, the mixture of the contaminant(s) and the sulfur mixture which is composed from 95% S and 5% aromatic/or aliphatic hydrocarbons used as polymerizing agents for sulfur (by weight), were added in a stainless steel vessel and primarily heated to 40 degree C for four hours, this time was sufficient for homogeneous mixing of the metals with sulfur and Na 2 S (to convert the metals to their corresponding sulfides). Additional SPC was then added and the temperature of the mixture was raised to 135 ±1 degree C, resulting in a molten form that was poured into a stainless steel mold where it cooled and solidified. Durability of the fabricated SPC matrices was assessed in terms of water of immersion, porosity, and compressive strength. The water absorption and open porosity were very low and didn't exceed 2.5 % for all matrices, whereas the compressive strength ranged between 7 and 14 KN/m 2 depending on the matrix composition. The immobilized waste forms of SPC were characterized by X-ray diffraction (XRD) technique that showed that the different contaminants were stabilized during the solidification process to form stable sulfides. Leachability of the waste matrices was assessed by the Toxicity Characteristic Leaching Procedure (TCLP) of the EPA, optimized and compared with the new EPA Universal Treatment Standards.The TCLP results showed that the concentration of the most contaminants released were under detection limit

Radioactive waste management and regulation

International Nuclear Information System (INIS)

Willrich, M.; Lester, R.K.; Greenberg, S.C.; Mitchell, H.C.; Walker, D.A.

1977-01-01

Purpose of this book is to assist in developing public policy and institutions for the safe management of radioactive waste, currently and long term. Both high-level waste and low-level waste containing transuranium elements are covered. The following conclusions are drawn: the safe management of post-fission radioactive waste is already a present necessity and an irreversible long-term commitment; the basic goals of U.S. radioactive waste policy are unclear; the existing organization for radioactive waste management is likely to be unworkable if left unchanged; and the existing framework for radioactive waste regulation is likely to be ineffective if left unchanged. The following recommendations are made: a national Radioactive Waste Authority should be established as a federally chartered public corporation; with NRC as the primary agency, a comprehensive regulatory framework should be established to assure the safety of all radioactive waste management operations under U.S. jurisdiction or control; ERDA should continue to have primary government responsibility for R and D and demonstration of radioactive waste technology; and the U.S. government should propose that an international Radioactive Waste Commission be established under the IAEA

Radioactive Waste Management Basis

International Nuclear Information System (INIS)

Perkins, B.K.

2009-01-01

The purpose of this Radioactive Waste Management Basis is to describe the systematic approach for planning, executing, and evaluating the management of radioactive waste at LLNL. The implementation of this document will ensure that waste management activities at LLNL are conducted in compliance with the requirements of DOE Order 435.1, Radioactive Waste Management, and the Implementation Guide for DOE Manual 435.1-1, Radioactive Waste Management Manual. Technical justification is provided where methods for meeting the requirements of DOE Order 435.1 deviate from the DOE Manual 435.1-1 and Implementation Guide.

Radioactive wastes and discharges

International Nuclear Information System (INIS)

1993-01-01

According to the Section 24 of the Finnish Radiation Decree (1512/91), the Finnish Centre for Radiation and Nuclear Safety shall specify the concentration and activity limits and principles for the determination whether a waste can be defined as a radioactive waste or not. The radiation safety requirements and limits for the disposal of radioactive waste are given in the guide. They must be observed when discharging radioactive waste into the atmosphere or sewer system, or when delivering solid low-activity waste to a landfill site without a separate waste disposal plan. The guide does not apply to the radioactive waste resulting from the utilization of nuclear energy of natural resources. (4 refs., 1 tab.)

Technology, socio-political acceptance, and the low-level radioactive waste disposal facility

International Nuclear Information System (INIS)

Andrews, L.J.; Domenech, J.S.

1986-01-01

The technology which is required to develop and operate low-level radioactive waste disposal sites in the 1990's is available today. The push for best available technology is a response to the political difficulties in securing public acceptance of the site selection process. Advances in waste management technologies include development of High Integrity Containers (HIC), solidification media, liquid volume reduction techniques using GEODE/sub sm/ and DeVoe-Holbein technology of selective removal of target radioisotopes, and CASTOR V storage casks. Advances in technology alone, however, do not make the site selection process easier and without socio-political acceptance there may be no process at all. Chem-Nuclear has been successful in achieving community acceptance at the Barnwell facility and elsewhere. For example, last June in Fall River County, South Dakota, citizens voted almost 2:1 to support the development of a low-level radioactive waste disposal facility. In Edgemont, the city nearest the proposed site, 85% of the voters were in favor of the proposed facility

Management of radioactive wastes

International Nuclear Information System (INIS)

Hendee, W.R.

1984-01-01

The disposal of radioactive wastes is perhaps the most controversial and least understood aspect of the use of nuclear materials in generating electrical power, the investigation of biochemical processes through tracer kinetics, and the diagnosis and treatment of disease. In the siting of nuclear power facilities, the disposal of radioactive wastes is invariably posed as the ultimate unanswerable question. In the fall of 1979, biochemical and physiologic research employing radioactive tracers was threatened with a slowdown resulting from temporary closure of sites for disposal of low-level radioactive wastes (LLW). Radioactive pharmaceuticals used extensively for diagnosis and treatment of human disease have increased dramatically in price, partly as a result of the escalating cost of disposing of radioactive wastes created during production of the labeled pharmaceuticals. These problems have resulted in identification of the disposal of LLW as the most pressing issue in the entire scheme of management of hazardous wastes. How this issue as well as the separate issue of disposal of high-level radioactive wastes (HLW) are being addressed at both national and state levels is the subject of this chapter

Radioactive waste disposal

International Nuclear Information System (INIS)

Bohm, H.; Closs, K.D.; Kuhn, K.

1981-01-01

The solutions to the technical problem of the disposal of radioactive waste are limited by a) the state of knowledge of reprocessing possibilites, b) public acceptance of the use of those techniques which are known, c) legislative procedures linking licensing of new nuclear power plants to the solution of waste problems, and d) other political constraints. Wastes are generated in the mining and enriching of radioactive elements, and in the operation of nuclear power plants as well as in all fields where radioactive substances may be used. Waste management will depend on the stability and concentration of radioactive materials which must be stored, and a resolution of the tension between numerous small storage sites and a few large ones, which again face problems of public acceptability

Effect of borate concentration on solidification of radioactive wastes by different cements

International Nuclear Information System (INIS)

Sun Qina; Li Junfeng; Wang Jianlong

2011-01-01

Highlights: → The effect of borate on cementation of radioactive borate evaporator concentrates by sulfoaluminate cement (SAC) and Portland cement (PC) was compared. → The X-ray diffraction (XRD) revealed that borate did not interfere with the formation of main hydration products of SAC and PC. → Borate, in the form of B(OH) 4- , incorporated in ettringite as solid solution phase. - Abstract: To investigate the effect of borate on the cementation of radioactive evaporator concentrates, and to provide more data for solidification formula optimization, the simulated borate evaporator concentrates with different borate concentrations (as B) and Na/B ratio (molar ratio) were solidified by sulfoaluminate cement (SAC) and Portland cement (PC), with addition of Ca(OH) 2 , zeolite and accelerator or water reducer. The hydration products of solidified matrices were characterized by X-ray diffraction (XRD). The experimental results showed that borate retarded the cement setting for both SAC and PC formulas, and the final setting time prolonged with decrease of Na/B ratio. Borate could enhance the fluidity of the cement mixture. The 28 d compressive strengths of the solidified matrices for both SAC and PC formulas decreased with increase of borate concentration. The XRD patterns suggested that, in the matrices maintained for 28 d, borate did not interfere with the formation of main hydration products of SAC and PC. Borate, in the form of B(OH) 4- , incorporated in ettringite (3CaO.Al 2 O 3 .3CaSO 4 .32H 2 O) as solid solution phase. The formula of SAC and PC developed in this study was effective for cementation of the simulated borate evaporator concentrates. However further optimization was required to reduce retarding effect of higher borate concentrations and to extend the practical feasibility for actual evaporator concentrates.

Radioactive wastes and discharges

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-07-01

The guide sets out the radiation safety requirements and limits for the treatment of radioactive waste. They shall be observed when discharging radioactive substances into the atmosphere or sewer system, or when delivering solid, low-activity waste to a landfill site without a separate waste treatment plan. The guide does not apply to the radioactive waste resulting from the utilisation of nuclear energy or natural resources.

Radioactive wastes and discharges

International Nuclear Information System (INIS)

2000-01-01

The guide sets out the radiation safety requirements and limits for the treatment of radioactive waste. They shall be observed when discharging radioactive substances into the atmosphere or sewer system, or when delivering solid, low-activity waste to a landfill site without a separate waste treatment plan. The guide does not apply to the radioactive waste resulting from the utilisation of nuclear energy or natural resources

International trends of radioactive waste management

International Nuclear Information System (INIS)

Luo Shanggeng

1989-01-01

The new trends of radioactive waste management in the world such as focusing on decreasing the amount of radioactive wastes, developing decontamination and decommissioning technology, conscientious solution for radiactive waste disposal, carrying out social services of waste treatment and quality assurance are reviewed. Besides, comments and suggestions are presented. Key words Radioactive waste management, Radioactive waste treatment, Radioactive waste disposal

Radioactive waste management

International Nuclear Information System (INIS)

2003-01-01

Almost all IAEA Member States use radioactive sources in medicine, industry, agriculture and scientific research, and countries remain responsible for the safe handling and storage of all radioactively contaminated waste that result from such activities. In some cases, waste must be specially treated or conditioned before storage and/or disposal. The Department of Technical Co-operation is sponsoring a programme with the support of the Nuclear Energy Department aimed at establishing appropriate technologies and procedures for managing radioactive wastes. (IAEA)

Objectives for radioactive waste packaging

International Nuclear Information System (INIS)

Flowers, R.H.

1982-04-01

The report falls under the headings: introduction; the nature of radioactive wastes; how to manage radioactive wastes; packaging of radioactive wastes (supervised storage; disposal); waste form evaluation and test requirements (supervised storage; disposal); conclusions. (U.K.)

Solidification of problem wastes: Annual progress report, October 1985-September 1986

International Nuclear Information System (INIS)

Franz, E.M.; Heiser, J.H. III; Colombo, P.

1987-02-01

This report describes initial work on the development of solidification systems for sodium nitrate waste and compacted waste. Sodium nitrate waste has been solidified in three types of materials: polyethylene, polyester-styrene (PES), and latex cement. Evaluations of the properties of the waste form, such as the ANS 16.1 leaching test, water immersion test and compressive strength measurements were performed on the waste forms containing various amounts of sodium nitrate. 9 refs., 9 figs., 7 tabs

Radioactive Wastes. Revised.

Science.gov (United States)

Fox, Charles H.

This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. This booklet deals with the handling, processing and disposal of radioactive wastes. Among the topics discussed are: The Nature of Radioactive Wastes; Waste Management; and Research and Development. There are…

Radioactive Waste Management Strategy

International Nuclear Information System (INIS)

2002-01-01

This strategy defines methods and means how collect, transport and bury radioactive waste safely. It includes low level radiation waste and high level radiation waste. In the strategy are foreseen main principles and ways of storage radioactive waste

Sulfur polymer stabilization/solidification (SPSS) treatment of mixed waste mercury recovered from environmental restoration activities at BNL

International Nuclear Information System (INIS)

Kalb, P.; Adams, J.; Milian, L.

2001-01-01

Over 1,140 yd 3 of radioactively contaminated soil containing toxic mercury (Hg) and several liters of mixed-waste elemental mercury were generated during a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) removal action at Brookhaven National Laboratory (BNL). The US Department of Energy's (DOE) Office of Science and Technology Mixed Waste Focus Area (DOE MWFA) is sponsoring a comparison of several technologies that may be used to treat these wastes and similar wastes at BNL and other sites across the DOE complex. This report describes work conducted at BNL on the application and pilot-scale demonstration of the newly developed Sulfur Polymer Stabilization/Solidification (SPSS) process for treatment of contaminated mixed-waste soils containing high concentrations (approximately 5,000 mg/L) of mercury and liquid elemental mercury. BNL's SPSS (patent pending) process chemically stabilizes the mercury to reduce vapor pressure and leachability and physically encapsulates the waste in a solid matrix to eliminate dispersion and provide long-term durability. Two 55-gallon drums of mixed-waste soil containing high concentrations of mercury and about 62 kg of radioactive contaminated elemental mercury were successfully treated. Waste loadings of 60 wt% soil were achieved without resulting in any increase in waste volume, while elemental mercury was solidified at a waste loading of 33 wt% mercury. Toxicity Characteristic Leaching Procedure (TCLP) analyses indicate the final waste form products pass current Environmental Protection Agency (EPA) allowable TCLP concentrations as well as the more stringent proposed Universal Treatment Standards. Mass balance measurements show that 99.7% of the mercury treated was successfully retained within the waste form, while only 0.3% was captured in the off gas system

Radioactive liquid waste filtering device

International Nuclear Information System (INIS)

Inami, Ichiro; Tabata, Masayuki; Kubo, Koji.

1988-01-01

Purpose: To prevent clogging in filter materials and improve the filtration performance for radioactive liquid wastes without increasing the amount of radioactive wastes. Constitution: In a radioactive waste filtering device, a liquid waste recycling pipe and a liquid recycling pump are disposed for recycling the radioactive liquid wastes in a liquid wastes vessel. In this case, the recycling pipe and the recycling pump are properly selected so as to satisfy the conditions capable of making the radioactive liquid wastes flowing through the pipe to have the Reynolds number of 10 4 - 10 5 . By repeating the transportation of radioactive liquid wastes in the liquid waste vessel through the liquid waste recycling pipe by the liquid waste recycling pump and then returning them to the liquid waste vessel again, particles of fine grain size in the suspended liquids are coagulated with each other upon collision to increase the grain size of the suspended particles. In this way, clogging of the filter materials caused by the particles of fine grain size can be prevented, thereby enabling to prevent the increase in the rising rate of the filtration differential pressure, reduce the frequency for the occurrence of radioactive wastes such as filter sludges and improve the processing performance. (Kamimura, M.)

Radioactive Waste Repositories Administration - SURAO

International Nuclear Information System (INIS)

Kucerka, M.

1998-01-01

The Atomic Act specifies, among other things, responsibilities of the government in the field of safe disposal of radioactive wastes. To satisfy this responsibility, the Ministry of Industry and Trade has established the Radioactive Waste Repositories Administration (SURAO). SURAO's major responsibilities include: (a) the preparation, construction, commissioning, operation, and decommissioning of radioactive waste repositories and the monitoring of their environmental impacts; (b) radioactive waste management; (c) spent or irradiated nuclear fuel processing into a form suitable for storage/disposal or reuse; (d) record-keeping of received radioactive wastes and their producers; (e) administration of fund transfers as stipulated by the Atomic Act, Article 27; (f) development of proposals for specification of fees to be paid to the Nuclear Account; (g) responsibility for and coordination of research and development in the field of radioactive waste handling and management; (h) supervision of licensees' margin earmarked for the decommissioning of their facilities; (i) providing services in radioactive waste handling and management; (j) handling and management of radioactive wastes that have been transferred to the Czech Republic from abroad and cannot be sent back; (k) interim administration of radioactive wastes that have become state property. The Statute of the Administration is reproduced in full. (P.A.)

Perspectives concerning radioactive waste management

International Nuclear Information System (INIS)

Noynaert, L.

2013-01-01

The article presents a general overview of the principles of radioactive waste management as established by the International Atomic Energy Agency. Subsequently, research and development related to radioactive waste management at the Belgian Nuclear Research Center SCK·CEN is discussed. Different topical areas are treated including radioactive waste characterisation, decontamination and the long-term management of radioactive waste. The decommissioning of the BR3 reactor and the construction and the exploitation of the underground research laboratory HADES are cited as examples of the pioneering role that SCK·CEN has played in radioactive waste management.

An interim report of the Subcommittee on Radioactive Waste Countermeasures: measures for radioactive waste treatment and disposal

International Nuclear Information System (INIS)

1984-01-01

The Subcommittee on Radioactive Waste Countermeasures has studied on the measures for land disposal of low-level radioactive wastes and ultra-low-level radioactive wastes and the measures for treatment and disposal of high-level radioactive wastes and transuranium wastes. The results of studies so far are presented as an interim report. In disposal of low-level radioactive wastes, the land disposal is being required increasingly. The measures according to the levels of radioactivity are necessary. For the ultra-low-level radioactive wastes, their occurrence in large quantities is expected along with reactor decommissioning. In disposal of the high-level radioactive wastes, the present status is a transition toward the practical stages. Transuranium wastes should increase in their arising in the future. (Mori, K.)

Cadarache LOR (liquides organiques radioactifs) treatment by a solidification process using NOCHAR polymers

International Nuclear Information System (INIS)

Vaudey, Claire-Emilie; Renou, Sebastien; Kelley, Dennis; Cochaud, Chantal; Serrano, Roger

2013-01-01

In France, two options can be considered to handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW). The first one is the incineration at CENTRACO facility and the second one is the disposal at ANDRA sites. The waste acceptance in these two channels is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the channel specifications. If the waste characteristics and the channel specifications (presence of significant quantities of halogens, complexing agents, organic components... or/and high activity limits) are incompatible, an alternative solution have to be identify. It consists of a waste pre-treatment process. For Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. They are composed of a mix of organic liquids and water: for the first one, 19 % of organic compounds (xylene, mesitylene, diphenyloxazole, TBP...) and 86.9 % of water, and for the second one, 23 % of organic compounds (TBP...) and 77 % of water. They contain halogens (chlorine and fluorine), complexants agents (nitrate, sulphate, oxalate and formate) and have got αβγ spectra with mass activities equal to some 100 Bq/g. Therefore, tritium is also present. As a consequence, in order for storage acceptance at the ANDRA site, it is necessary to pre-treat the waste. An adequate solution seems to be a solidification process using NOCHAR polymers. Indeed, NOCHAR polymers correspond to an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing ...) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and

Waste minimization for commercial radioactive materials users generating low-level radioactive waste

International Nuclear Information System (INIS)

Fischer, D.K.; Gitt, M.; Williams, G.A.; Branch, S.; Otis, M.D.; McKenzie-Carter, M.A.; Schurman, D.L.

1991-07-01

The objective of this document is to provide a resource for all states and compact regions interested in promoting the minimization of low-level radioactive waste (LLW). This project was initiated by the Commonwealth of Massachusetts, and Massachusetts waste streams have been used as examples; however, the methods of analysis presented here are applicable to similar waste streams generated elsewhere. This document is a guide for states/compact regions to use in developing a system to evaluate and prioritize various waste minimization techniques in order to encourage individual radioactive materials users (LLW generators) to consider these techniques in their own independent evaluations. This review discusses the application of specific waste minimization techniques to waste streams characteristic of three categories of radioactive materials users: (1) industrial operations using radioactive materials in the manufacture of commercial products, (2) health care institutions, including hospitals and clinics, and (3) educational and research institutions. Massachusetts waste stream characterization data from key radioactive materials users in each category are used to illustrate the applicability of various minimization techniques. The utility group is not included because extensive information specific to this category of LLW generators is available in the literature

Radioactive mixed waste disposal

International Nuclear Information System (INIS)

Jasen, W.G.; Erpenbeck, E.G.

1993-02-01

Various types of waste have been generated during the 50-year history of the Hanford Site. Regulatory changes in the last 20 years have provided the emphasis for better management of these wastes. Interpretations of the Atomic Energy Act of 1954 (AEA), the Resource Conservation and Recovery Act of 1976 (RCRA), and the Hazardous and Solid Waste Amendments (HSWA) have led to the definition of radioactive mixed wastes (RMW). The radioactive and hazardous properties of these wastes have resulted in the initiation of special projects for the management of these wastes. Other solid wastes at the Hanford Site include low-level wastes, transuranic (TRU), and nonradioactive hazardous wastes. This paper describes a system for the treatment, storage, and disposal (TSD) of solid radioactive waste

Radioactive waste management

International Nuclear Information System (INIS)

1984-07-01

The purpose of this document is to set out the Government's current strategy for the long term in the management of radioactive wastes. It takes account of the latest developments, and will be subject to review in the light of future developments and studies. The subject is discussed under the headings: what are radioactive wastes; who is responsible; what monitoring takes place; disposal as the objective; low-level wastes; intermediate-level wastes; discharges from Sellafield; heat generating wastes; how will waste management systems and procedures be assessed; how much more waste is there going to be in future; conclusion. (U.K.)

Radioactive waste (disposal)

International Nuclear Information System (INIS)

Jenkin, P.

1985-01-01

The disposal of low- and intermediate-level radioactive wastes was discussed. The following aspects were covered: public consultation on the principles for assessing disposal facilities; procedures for dealing with the possible sites which the Nuclear Industry Radioactive Waste Executive (NIREX) had originally identified; geological investigations to be carried out by NIREX to search for alternative sites; announcement that proposal for a site at Billingham is not to proceed further; NIREX membership; storage of radioactive wastes; public inquiries; social and environmental aspects; safety aspects; interest groups; public relations; government policies. (U.K.)

Method of treating the waste liquid of a washing containing a radioactive substance

International Nuclear Information System (INIS)

Sawaguchi, Yusuke; Tsuyuki, Takashi; Kaneko, Masato; Sato, Yasuhiko; Yamaguchi, Takashi.

1975-01-01

Object: To separate waste liquid resulting from washing and which contains a radioactive substance and surface active agent into high purity water and a solid waste substance containing a small quantity of surface active agent. Structure: To waste liquid from a waste liquid tank is added a pH adjusting agent for adjusting the pH to 5.5, and the resultant liquid is sent to an agglomeration reaction tank, in which an inorganic agglomerating agent is added to the waste liquid to cause a major proportion of the radioactive substance and surface active agent to form flocks produced through agglomeration. Then, the waste liquid is sent from the agglomeration reaction tank to a froth separation tank, to which air is supplied through a perforated plate to cause frothing. The over-flowing liquid is de-frothed, and then the insoluble matter is separated as sludge, followed by hydroextraction and drying for solidification. The treated liquid extracted from a froth separation tank is sent to an agglomerating agent recovery tank for separation of the agglomeration agent, and then the residual surface active agent is removed by adsorption in an active carbon adsorption tower, followed by concentration by evaporation in an evaporating can. The concentrated liquid is extracted and then solidified with cement or asphalt. (Kamimura, M.)

Radioactive wastes

International Nuclear Information System (INIS)

Dupuis, M.C.

2007-01-01

Managing radioactive wastes used to be a peripheral activity for the French atomic energy commission (Cea). Over the past 40 years, it has become a full-fledged phase in the fuel cycle of producing electricity from the atom. In 2005, the national radioactive waste management agency (ANDRA) presented to the government a comprehensive overview of the results drawn from 15 years of research. This landmark report has received recognition beyond France's borders. By broadening this agency's powers, an act of 28 June 2006 acknowledges the progress made and the quality of the results. It also sets an objective for the coming years: work out solutions for managing all forms of radioactive wastes. The possibility of recovering wastes packages from the disposal site must be assured as it was asked by the government in 1998. The next step will be the official demand for the creation of a geological disposal site in 2016

What to do with radioactive wastes?

International Nuclear Information System (INIS)

2006-01-01

This power point presentation (82 slides) gives information on what is a radioactive waste, radioactivity and historical review of radioactivity, radioactive period, natural radioactivity (with examples of data), the three main radiation types (α, β, γ), the origin of radioactive wastes (nuclear power, research, defense, other), the proportion of radioactive wastes in the total of industrial wastes in France, the classification of nuclear wastes according to their activity and period, the quantities and their storage means, the 1991 december 30 law (France) related to the radioactive waste management, the situation in other countries (Germany, Belgium, Canada, USA, Finland, Japan, Netherlands, Sweden, Switzerland), volume figures and previsions for the various waste types in 2004, 2010 and 2020, the storage perspectives, the French national debate on radioactive waste management and the objective of perpetuated solutions, the enhancement of the public information, the 15 June 2006 law on a sustainable management of radioactive materials and wastes with three main axis (deep separation and transmutation, deep storage, waste conditioning and long term surface storage), and the development of a nuclear safety and waste culture that could be extended to other types of industry

The disposal of radioactive waste

International Nuclear Information System (INIS)

Ormai, P.

2006-01-01

The first part shows different ways of 'producing' radioactive wastes, defines the wastes of small, medium and high activity and gives estimation on the quantity of the necessary capacities of waste disposal facilities. The modern radioactive waste disposal that is the integrated processing of the form of waste, the package, the technical facility and the embedding geological environment that guarantee the isolation together. Another factor is the lifetime of radioactive waste which means that any waste containing long lifetime waste in higher concentration than 400-4000 kBq/kg should be disposed geologically. Today the centre of debate disposal of radioactive waste is more social than technical. For this reason not only geological conditions and technical preparations, but social discussions and accepting communities are needed in selecting place of facilities. Now, the focus is on long term temporary disposal of high activity wastes, like burnt out heating elements. The final part of the paper summarizes the current Hungarian situation of disposal of radioactive wastes. (T-R.A.)

Radioactive waste processing

International Nuclear Information System (INIS)

Curtiss, D.H.; Heacock, H.W.

1976-01-01

The description is given of a process for treating radioactive waste whereby a mud of radioactive waste and cementing material is formed in a mixer. This mud is then transferred from the mixer to a storage and transport container where it is allowed to harden. To improve transport efficiency an alkali silicate or an alkaline-earth metal silicate is added to the mud. For one hundred parts by weight of radioactive waste in the mud, twenty to one hundred parts by weight of cementing material are added and five to fifty parts by weight of silicate, the amount of waste in the mud exceeding the combined amount of cementing and silicate material [fr

Properties and solidification of decontamination wastes

International Nuclear Information System (INIS)

Davis, M.S.; Piciulo, P.L.; Bowerman, B.S.; Adams, J.W.; Milian, L.

1983-01-01

LWRs will require one or more chemical decontaminations to achieve their designed lifetimes. Primary system decontamination is designed to lower radiation fields in areas where plant maintenance personnel must work. Chemical decontamination methods are either hard (concentrated chemicals, approximately 5 to 25 weight percent) or soft (dilute chemicals less than 1 percent by weight). These methods may have different chemical reagents, some tailor-made to the crud composition and many methods are and will be proprietary. One factor common to most commercially available processes is the presence of organic acids and chelates. These types of organic reagents are known to enhance the migration of radionuclides after disposal in a shallow land burial site. The NRC sponsors two programs at Brookhaven National Laboratory that are concerned with the management of decontamination wastes which will be generated by the full system decontamination of LWRs. These two programs focus on potential methods for degrading or converting decontamination wastes to more acceptable forms prior to disposal and the impact of disposing of solidified decontamination wastes. The results of the solidification of simulated decontamination resin wastes will be presented. Recent results on combustion of simulated decontamintion wastes will be described and procedures for evaluating the release of decontamination reagents from solidified wastes will be summarized

Development of an integrated liquid radioactive waste management system. Part of a coordinated programme on integrated radioactive waste management systems and their impact on the environment

International Nuclear Information System (INIS)

Pavlik, O.

1976-09-01

This final report discusses the theory of the air heated non-boiling evaporation and the laboratory and pilot plant experiences of simultaneous evaporation and solidification. The evaporator concentrates and dries the radioactive liquid wastes by hot air-flow circulating in closed circuit. The air is used both for heating and removal of vapours. The liquid waste becomes saturated and the salt cake is accumulated in the tank. The yield evaporation rate and the decontamination factor or 90 Sr and 137 Cs was investigated as a function to air-flow, air temperature and humidity. During the evaporation process a simultaneous cementation method was developed to fix the water soluble components of the salt cake. The leaching rate of 90 Sr and 137 Cs, the firmness and the homogeneity of the concrete was investigated

Radioactive Waste Management BasisApril 2006

Energy Technology Data Exchange (ETDEWEB)

Perkins, B K

2011-08-31

This Radioactive Waste Management Basis (RWMB) documents radioactive waste management practices adopted at Lawrence Livermore National Laboratory (LLNL) pursuant to Department of Energy (DOE) Order 435.1, Radioactive Waste Management. The purpose of this Radioactive Waste Management Basis is to describe the systematic approach for planning, executing, and evaluating the management of radioactive waste at LLNL. The implementation of this document will ensure that waste management activities at LLNL are conducted in compliance with the requirements of DOE Order 435.1, Radioactive Waste Management, and the Implementation Guide for DOE Manual 435.1-1, Radioactive Waste Management Manual. Technical justification is provided where methods for meeting the requirements of DOE Order 435.1 deviate from the DOE Manual 435.1-1 and Implementation Guide.

Stabilization and Solidification of Nitric Acid Effluent Waste at Y-12

Energy Technology Data Exchange (ETDEWEB)

Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); Lorenzo-Martin, Cinta [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-16

Consolidated Nuclear Security, LLC (CNS) at the Y-12 plant is investigating approaches for the treatment (stabilization and solidification) of a nitric acid waste effluent that contains uranium. Because the pH of the waste stream is 1-2, it is a difficult waste stream to treat and stabilize by a standard cement-based process. Alternative waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the nitric acid effluent wastes.

Radioactive waste management and regulation

International Nuclear Information System (INIS)

Willrich, M.

1976-12-01

The following conclusions are reached: (1) safe management of post-fission radioactive waste is already a present necessity and an irreversible long-term commitment; (2) basic goals of U.S. radioactive waste policy are unclear; (3) the existing organization for radioactive waste management is likely to be unworkable if left unchanged; and (4) the existing framework for radioactive waste regulation is likely to be ineffective if left unchanged

Radioactive waste management

International Nuclear Information System (INIS)

Morley, F.

1980-01-01

A summary is given of the report of an Expert Group appointed in 1976 to consider the 1959 White Paper 'The Control of Radioactive Wastes' in the light of the changes that have taken place since it was written and with the extended remit of examining 'waste management' rather than the original 'waste disposal'. The Group undertook to; review the categories and quantities present and future of radioactive wastes, recommend the principles for the proper management of these wastes, advise whether any changes in practice or statutory controls are necessary and make recommendations. (UK)

Method of manufacturing borosilicate glass solidification products

International Nuclear Information System (INIS)

Tanaka, Tsuneya.

1986-01-01

Purpose: To obtain glass solidification products efficiently in a dry process from medium and high level radioactive liquid wastes discharged from PWR type reactors. Method: Boric acid-containing radioactive liquid wastes generated from primary coolants of PWR type reactors are evaporated to condensate as the pre-treatment. The concentrated liquid wastes are supplied to a drum type rotary kiln. While on the other hand, usual glass frits are introduced into the kiln. The liquid wastes are dried in the rotary kiln, as well as B 2 O 3 and the glass frits in the liquid wastes are combined into glass particles. In this case, since the kiln is rotated, no glass particles are deposited on the wall of the kiln. Then, the glass particles are introduced for melting into a high frequency melting furnace made of metal. The melting temperature is set to 1100 - 1150 deg C. The molten borosilicate glass is recovered from the bottom of the melting furance, contained in a canister and cooled for several hours, and then a cover is welded to the canister. (Ikeda, J.)

Management of radioactive wastes from nuclear fuels and power plants in Canada

International Nuclear Information System (INIS)

Tomlinson, M.; Mayman, S.A.; Tammemagi, H.Y.; Gale, J.; Sanford, B.; Dyne, P.J.

1977-01-01

The nature of Canadian nuclear fuel and nuclear generating plant radioactive wastes are summarized. Full exploitation of fission energy resources entails recovery of all fissile and fertile material from spent fuel and separating the fission products as wastes for disposal. A plan for final disposal of all the radioactive wastes is a key component of the waste management scheme. Principles of a scheme for safe, responsible disposal of long-lived radioactive wastes deep underground in isolation from man and the biosphere are outlined. The status of the development and construction program is indicated. We plan to select a site in either a hard rock formation or in a suitable salt bed by 1981 so that a repository can be constructed to begin a demonstration phase in 1986. The repository is to be capable of eventual expansion to accomodate all Canadian nuclear wastes to at least 2050 when in full-scale operation. Extensive geotechnical studies have been initiated in order to select a site, and design and test the repository. We have demonstrated incorporation of fission products in solids that in the short term (17 years) dissolve more slowly than plutonium decays. Investigations of long-term stability are in hand. The principle of retardation of migration of fission products, so that they decay before surfacing, has been tested. Additional capacity for storage of used fuel prior to reprocessing and disposal is required by 1986 and a preliminary design has been prepared for a pool facility to be located at a central fuel recycling and disposal complex. A demonstration of dry storage of fuel in concrete containers is in progress. The quantities of CANDU generating-station wastes and the principles and methods for managing them are summarized. Methods for volume reduction and immobilization by solidification are well advanced. A radioactive-waste operations site is being developed with several different types of surface storage, each with multiple barriers against

Management of radioactive wastes from nuclear fuels and power plants in Canada

International Nuclear Information System (INIS)

Tomlinson, M.; Mayman, S.A.; Tammemagi, H.Y.; Gale, J.; Sanford, B.

1977-01-01

The nature of Canadian nuclear fuel and nuclear generating plant radioactive wastes is summarized. Full exploitation of fission energy resources entails recovery of all fissile and fertile material from spent fuel and separating the fission products as wastes for disposal. A plan for final disposal of all the radioactive wastes is a key component of the waste management scheme. Principles of a scheme for safe, responsible disposal of long-lived radioactive wastes deep underground, in isolation from man and the biosphere, are outlined. The status of the development and construction programme is indicated. It is planned to select a site in either a hard rock formation or in a suitable salt bed by 1981 so that a repository can be constructed to begin a demonstration phase in 1986. The repository is to be capable of eventual expansion to accomodate all Canadian nuclear wastes to at least 2050 when in full-scale operation. Extensive geotechnical studies have been initiated in order to select a site, and design and test the repository. The incorporation of fission products in solids that in the short term (17 years) dissolve more slowly than plutonium decays has been demonstrated. Investigations of long-term stability are in hand. The principle of retardation of migration of fission products, so that they decay before surfacing, has been tested. Additional capacity for storage of used fuel prior to reprocessing and disposal is required by 1986 and a preliminary design has been prepared for a pool facility to be located at a central fuel recycling and disposal complex. A demonstration of dry storage of fuel in concrete containers is in progress. The quantities of CANDU generating-station wastes and the principles and methods for managing them are summarized. Methods for volume reduction and immobilization by solidification are well advanced. A radioactive-waste operations site is being developed with several different types of surface storage, each with multiple barriers

Predisposal Radioactive Waste Management

International Nuclear Information System (INIS)

2014-01-01

Recognition of the importance of the safe management of radioactive waste means that, over the years, many well-established and effective techniques have been developed, and the nuclear industry and governments have gained considerable experience in this field. Minimization of waste is a fundamental principle underpinning the design and operation of all nuclear operations, together with waste reuse and recycling. For the remaining radioactive waste that will be produced, it is essential that there is a well defined plan (called a waste treatment path) to ensure the safe management and ultimately the safe disposal of radioactive waste so as to guarantee the sustainable long term deployment of nuclear technologies

Management of radioactive waste: A review

OpenAIRE

Luis Paulo Sant'ana; Taynara Cristina Cordeiro

2016-01-01

The issue of disposal of radioactive waste around the world is not solved by now and the principal reason is the lack of an efficient technologic system. The fact that radioactive waste decays of radioactivity with time are the main reasons for setting nuclear or radioactive waste apart from the other common hazardous wastes management. Radioactive waste can be classified according to the state of matter and level of radioactivity and this classification can be differently interpreted from co...

Study of integral waste management systems and their effects on the environment in Thailand. Part of a coordinated programme on migration and disperios of radionuclides from storage of radioactive waste under various conditions in the terrestrial environment

International Nuclear Information System (INIS)

Karasuddhi, P.

1976-02-01

The research work carried out is divided into 3 parts: 1. Experimental study of sorption capacity of Sr-90, Cs-137 and radioactive liquid waste of OAEP onto soil and clay. 2. The fixation of radioactive sludge and resin in cement, gumcrete and bitumen. The results show that the sludge from the OAEP waste treatment plant can be fixed very well with concrete, gumcrete and bitumen, but the solidification of resin in concrete, gumcrete and bitumen is not a good method for waste treatment. In the third part, the results of environment studies around the nuclear research center are presented

Solidification/stabilization of technetium in cement-based grouts

International Nuclear Information System (INIS)

Gilliam, T.M.; Bostick, W.D.; Spence, R.D.; Shoemaker, J.L.

1990-01-01

Mixed low-level radioactive and chemically hazardous process treatment wastes from the Portsmouth Gaseous Diffusion Plant are stabilized by solidification in cement-based grouts. Conventional portland cement and fly ash grouts have been shown to be effective for retention of hydrolyzable metals (e.g., lead, cadmium, uranium and nickel) but are marginally acceptable for retention of radioactive Tc-99, which is present in the waste as the highly mobile pertechnate anion. Addition of ground blast furnace slag to the grout is shown to reduce the leachability of technetium by several orders of magnitude. The selective effect of slag is believed to be due to its ability to reduce Tc(VII) to the less soluble Tc(IV) species. 12 refs., 4 tabs

Method of storing radioactive wastes

International Nuclear Information System (INIS)

Adachi, Toshio; Hiratake, Susumu.

1980-01-01

Purpose: To reduce the radiation doses externally irradiated from treated radioactive waste and also reduce the separation of radioactive nuclide due to external environmental factors such as air, water or the like. Method: Radioactive waste adhered with radioactive nuclide to solid material is molten to mix and submerge the radioactive nuclide adhered to the surface of the solid material into molten material. Then, the radioactive nuclide thus mixed is solidified to store the waste in solidified state. (Aizawa, K.)

Regulation of radioactive waste management

International Nuclear Information System (INIS)

2002-01-01

This bulletin contains information about activities of the Nuclear Regulatory Authority of the Slovak Republic (UJD). In this leaflet the regulation of radioactive waste management of the UJD are presented. Radioactive waste (RAW) is the gaseous, liquid or solid material that contains or is contaminated with radionuclides at concentrations or activities greater than clearance levels and for which no use is foreseen. The classification of radioactive waste on the basis of type and activity level is: - transition waste; - short lived low and intermediate level waste (LlLW-SL); - long lived low and intermediate level waste (LlLW-LL); - high level waste. Waste management (in accordance with Act 130/98 Coll.) involves collection, sorting, treatment, conditioning, transport and disposal of radioactive waste originated by nuclear facilities and conditioning, transport to repository and disposal of other radioactive waste (originated during medical, research and industrial use of radioactive sources). The final goal of radioactive waste management is RAW isolation using a system of engineered and natural barriers to protect population and environment. Nuclear Regulatory Authority of the Slovak Republic regulates radioactive waste management in accordance with Act 130/98 Coll. Inspectors regularly inspect and evaluate how the requirements for nuclear safety at nuclear facilities are fulfilled. On the basis of safety documentation evaluation, UJD issued permission for operation of four radioactive waste management facilities. Nuclear facility 'Technologies for treatment and conditioning contains bituminization plants and Bohunice conditioning centre with sorting, fragmentation, evaporation, incineration, supercompaction and cementation. Final product is waste package (Fibre reinforced container with solidified waste) acceptable for near surface repository in Mochovce. Republic repository in Mochovce is built for disposal of short lived low and intermediate level waste. Next

Sulfur polymer stabilization/solidification (SPSS) treatment of mixed waste mercury recovered from environmental restoration activities at BNL

Energy Technology Data Exchange (ETDEWEB)

Kalb, P.; Adams, J.; Milian, L.

2001-01-29

Over 1,140 yd{sup 3} of radioactively contaminated soil containing toxic mercury (Hg) and several liters of mixed-waste elemental mercury were generated during a Comprehensive Environmental Response, Compensation and Liability Act (CERCLA) removal action at Brookhaven National Laboratory (BNL). The US Department of Energy's (DOE) Office of Science and Technology Mixed Waste Focus Area (DOE MWFA) is sponsoring a comparison of several technologies that may be used to treat these wastes and similar wastes at BNL and other sites across the DOE complex. This report describes work conducted at BNL on the application and pilot-scale demonstration of the newly developed Sulfur Polymer Stabilization/Solidification (SPSS) process for treatment of contaminated mixed-waste soils containing high concentrations ({approximately} 5,000 mg/L) of mercury and liquid elemental mercury. BNL's SPSS (patent pending) process chemically stabilizes the mercury to reduce vapor pressure and leachability and physically encapsulates the waste in a solid matrix to eliminate dispersion and provide long-term durability. Two 55-gallon drums of mixed-waste soil containing high concentrations of mercury and about 62 kg of radioactive contaminated elemental mercury were successfully treated. Waste loadings of 60 wt% soil were achieved without resulting in any increase in waste volume, while elemental mercury was solidified at a waste loading of 33 wt% mercury. Toxicity Characteristic Leaching Procedure (TCLP) analyses indicate the final waste form products pass current Environmental Protection Agency (EPA) allowable TCLP concentrations as well as the more stringent proposed Universal Treatment Standards. Mass balance measurements show that 99.7% of the mercury treated was successfully retained within the waste form, while only 0.3% was captured in the off gas system.

Radioactive waste in Federal Germany

International Nuclear Information System (INIS)

Brennecke, P.; Schumacher, J.; Warnecke, E.

1988-01-01

The Physikalisch-Technische Bundesanstalt (PTB) is responsible for the long-term storage and disposal of radioactive waste according to the Federal Atomic Energy Act. On behalf of the Federal Minister of the Environment, Nature Conservation and Nuclear Safety, since 1985, the PTB has been carrying out annual inquiries into the amounts of radioactive waste produced in the Federal Republic of Germany. Within the scope of this inquiry performed for the preceding year, the amounts of unconditioned and conditioned waste are compiled on a producer- and plant-specific basis. On the basis of the inquiry for 1986 and of data presented to the PTB by the waste producers, future amounts of radioactive waste have been estimated up to the year 2000. The result of this forecast is presented. In the Federal Republic of Germany two sites are under consideration for disposal of radioactive waste. In the abandoned Konrad iron mine in Salzgitter-Bleckenstedt it is intended to dispose of such radioactive waste which has a negligible thermal influence upon the host rock. The Gorleben salt dome is being investigated for its suitability for the disposal of all kinds of solid and solidified radioactive wastes, especially of heat-generating waste. Comparing the estimated amount of radioactive wastes with the capacity of both repositories it may be concluded that the Konrad and Gorleben repositories will provide sufficient capacity to ensure the disposal of all kinds of radioactive waste on a long-term basis in the Federal Republic of Germany. 1 fig., 2 tabs

Method for calcining radioactive wastes

International Nuclear Information System (INIS)

Bjorklund, W.J.; McElroy, J.L.; Mendel, J.E.

1979-01-01

A method for the preparation of radioactive wastes in a low leachability form involves calcining the radioactive waste on a fluidized bed of glass frit, removing the calcined waste to melter to form a homogeneous melt of the glass and the calcined waste, and then solidifying the melt to encapsulate the radioactive calcine in a glass matrix

Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

International Nuclear Information System (INIS)

Knox, C.A.; Farnsworth, R.K.

1981-08-01

The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In this facility, which is installed in a radiochemical cell, small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. Initial operations were completed with nonradioactive, simulated waste solutions (Knox, Siemens and Berger 1981). The first radioactive operations in this facility were performed with a simulated, commercial waste composition containing tracer levels of 99 Tc and 131 I. This report describes the facility and test operations and presents the results of the behavior of 131 I and 99 Tc during solidification of radioactive liquid wastes. During the spray calcination of commercial high-level liquid waste spiked with 99 Tc and 131 I, there was a 0.3 wt% loss of particulates, a 0.15 wt% loss of 99 Tc and a 31 wt% loss of 131 I past the sintered-metal filters. These filters and a venturi scrubber were very efficient in removing particulates and 99 Tc from the off-gas stream. Liquid scrubbers were not efficient in removing 131 I, as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents will be needed to remove iodine. For all future RLSWTF operations where iodine is present, a silver zeolite adsorber will be used

ORNL radioactive waste operations

International Nuclear Information System (INIS)

Sease, J.D.; King, E.M.; Coobs, J.H.; Row, T.H.

1982-01-01

Since its beginning in 1943, ORNL has generated large amounts of solid, liquid, and gaseous radioactive waste material as a by-product of the basic research and development work carried out at the laboratory. The waste system at ORNL has been continually modified and updated to keep pace with the changing release requirements for radioactive wastes. Major upgrading projects are currently in progress. The operating record of ORNL waste operation has been excellent over many years. Recent surveillance of radioactivity in the Oak Ridge environs indicates that atmospheric concentrations of radioactivity were not significantly different from other areas in East Tennesseee. Concentrations of radioactivity in the Clinch River and in fish collected from the river were less than 4% of the permissible concentration and intake guides for individuals in the offsite environment. While some radioactivity was released to the environment from plant operations, the concentrations in all of the media sampled were well below established standards